PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

DOEpatents

Puechl, K.H.

1963-09-24

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

Cascading pressure reactor and method for solar-thermochemical reactions

DOEpatents

Ermanoski, Ivan

2017-11-14

Reactors and methods for solar thermochemical reactions are disclosed. The reactors and methods include a cascade of reduction chambers at successively lower pressures that leads to over an order of magnitude pressure decrease compared to a single-chambered design. The resulting efficiency gains are substantial, and represent an important step toward practical and efficient solar fuel production on a large scale.

Nuclear reactor pressure vessel support system

DOEpatents

Sepelak, George R.

1978-01-01

A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

U.S.-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program

NASA Astrophysics Data System (ADS)

Dabrowski, Richard S.

2014-08-01

The TOPAZ International Program (TIP) was the final name given to a series of projects to purchase and test the TOPAZ-II, a space-based nuclear reactor of a type that had been further developed in the Soviet Union than in the United States. In the changing political situation associated with the break-up of the Soviet Union it became possible for the United States to not just purchase the system, but also to employ Russian scientists, engineers and testing facilities to verify its reliability. The lessons learned from the TIP illuminate some of the institutional and cultural challenges to U.S. - Russian cooperation in technology research which remain true today.

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

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

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

1986-01-01

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

The current state of the Russian reduced enrichment research reactors program

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

Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

1997-08-01

During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% frommore » RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.« less

A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy

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

Beach, N. J.; Knapp, S. M. M.; Landis, C. R., E-mail: landis@chem.wisc.edu

The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from −90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor tomore » be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.« less

Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

NASA Astrophysics Data System (ADS)

Ivanov, Yu. A.

2007-12-01

An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.

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.

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

NASA Astrophysics Data System (ADS)

Kapychev, V.; Davydov, D.; Gorokhov, V.; Ioltukhovskiy, A.; Kazennov, Yu; Tebus, V.; Frolov, V.; Shikov, A.; Shishkov, N.; Kovalenko, V.; Shishkin, N.; Strebkov, Yu

2000-12-01

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water-graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm2) are considered. At the present time, development and fabrication of lithium orthosilicate-beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350-700°C. Technical problems associated with manufacturing of the modules are discussed.

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.

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.

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

Results of operation and current safety performance of nuclear facilities located in the Russian Federation

NASA Astrophysics Data System (ADS)

Kuznetsov, V. M.; Khvostova, M. S.

2016-12-01

After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104-109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation "Rosatom", should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn't happen. The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP

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

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…

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...

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

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

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

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required tomore » complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.« 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.

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.

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%).

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.

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

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

Trtilek, Radek; Podlaha, Josef

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the secondmore » shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)« less

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)

Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

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

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-03-01

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Returnmore » Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.« less

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

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.

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

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

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial

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

Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation

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

Svitak, F.; Broz, V.; Hrehor, M.

2008-07-15

The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF formore » transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)« less

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)

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

Russian EVA 34

NASA Image and Video Library

2013-08-16

ISS036-E-033400 (16 Aug. 2013) --- Russian cosmonaut Alexander Misurkin (lower left), Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the seven-hour, 29-minute spacewalk ? the longest ever conducted by a pair of Russian cosmonauts ? Misurkin and Fyodor Yurchikhin (out of frame) rigged cables for the future arrival of a Russian laboratory module and installed an experiment panel.

Russian EVA 34

NASA Image and Video Library

2013-08-16

ISS036-E-033402 (16 Aug. 2013) --- Russian cosmonaut Alexander Misurkin (lower left), Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the seven-hour, 29-minute spacewalk ? the longest ever conducted by a pair of Russian cosmonauts ? Misurkin and Fyodor Yurchikhin (out of frame) rigged cables for the future arrival of a Russian laboratory module and installed an experiment panel.

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.

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

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

Long-term storage facility for reactor compartments in Sayda Bay - German support for utilization of nuclear submarines in Russia

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

Wolff, Dietmar; Voelzke, Holger; Weber, Wolfgang

2007-07-01

The German-Russian project that is part of the G8 initiative on Global Partnership Against the Spread of Weapons and Materials of Mass Destruction focuses on the speedy construction of a land-based interim storage facility for nuclear submarine reactor compartments at Sayda Bay near Murmansk. This project includes the required infrastructure facilities for long-term storage of about 150 reactor compartments for a period of about 70 years. The interim storage facility is a precondition for effective activities of decommissioning and dismantlement of almost all nuclear-powered submarines of the Russian Northern Fleet. The project also includes the establishment of a computer-assisted wastemore » monitoring system. In addition, the project involves clearing Sayda Bay of other shipwrecks of the Russian navy. On the German side the project is carried out by the Energiewerke Nord GmbH (EWN) on behalf of the Federal Ministry of Economics and Labour (BMWi). On the Russian side the Kurchatov Institute holds the project management of the long-term interim storage facility in Sayda Bay, whilst the Nerpa Shipyard, which is about 25 km away from the storage facility, is dismantling the submarines and preparing the reactor compartments for long-term interim storage. The technical monitoring of the German part of this project, being implemented by BMWi, is the responsibility of the Federal Institute for Materials Research and Testing (BAM). This paper gives an overview of the German-Russian project and a brief description of solutions for nuclear submarine disposal in other countries. At Nerpa shipyard, being refurbished with logistic and technical support from Germany, the reactor compartments are sealed by welding, provided with biological shielding, subjected to surface treatment and conservation measures. Using floating docks, a tugboat tows the reactor compartments from Nerpa shipyard to the interim storage facility at Sayda Bay where they will be left on the on

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.

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.

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028082 (9 Nov. 2013) --- Russian cosmonaut Sergey Ryazanskiy, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Ryazanskiy and Russian cosmonaut Oleg Kotov (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028067 (9 Nov. 2013) --- Russian cosmonaut Oleg Kotov, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Russian cosmonaut Sergey Ryazanskiy (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028101 (9 Nov. 2013) --- Russian cosmonaut Oleg Kotov, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Russian cosmonaut Sergey Ryazanskiy (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028094 (9 Nov. 2013) --- Russian cosmonaut Oleg Kotov, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Russian cosmonaut Sergey Ryazanskiy (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028107 (9 Nov. 2013) --- Russian cosmonaut Oleg Kotov, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Russian cosmonaut Sergey Ryazanskiy (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028102 (9 Nov. 2013) --- Russian cosmonaut Oleg Kotov, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Russian cosmonaut Sergey Ryazanskiy (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Research on Russian National Character

ERIC Educational Resources Information Center

Na, Zhuo

2008-01-01

The special geographical location Russia lies in creates the unique character of the Russian nation. Based on the dual nature of the Russian national character, the Russian geographical environment and the analysis of its social structure, this text tries to explore the reasons of the dual nature of Russian national character.

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.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035204 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035130 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035129 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035124 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035133 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035205 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035126 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035163 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers.

Russian EVA 28

NASA Image and Video Library

2011-02-16

ISS026-E-027391 (16 Feb. 2011) --- Russian cosmonaut Dmitry Kondratyev, Expedition 26 flight engineer, wearing a Russian Orlan-MK spacesuit, participates in a session of extravehicular activity (EVA) focused on the installation of two scientific experiments outside the Zvezda Service Module of the International Space Station. During the four-hour, 51-minute spacewalk, Kondratyev and Russian cosmonaut Oleg Skripochka (out of frame), flight engineer, installed a pair of earthquake and lightning sensing experiments and retrieved a pair of spacecraft material evaluation panels.

Russian EVA 36

NASA Image and Video Library

2013-11-09

ISS037-E-028569 (9 Nov. 2013) --- Russian cosmonaut Oleg Kotov, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, uses a still camera during a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Russian cosmonaut Sergey Ryazanskiy (out of frame) continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

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.

Russian Flight Control Room

NASA Image and Video Library

2004-04-20

NASA Deputy Administrator Fred Gregory, left, joins Russian Federal Space Agency Deputy General-Director Nikolai Moiseev, Wednesday, April 21, 2004, at the Russian Mission Control Center outside Moscow to view the docking of the Expedition 9 crew to the International Space Station in a Russian Soyuz spacecraft. Photo Credit: (NASA/Bill Ingalls)

Fire safety arrangement of inhabited pressurized compartments of manned spacecraft

NASA Astrophysics Data System (ADS)

Bolodian, Ivan; Melikhov, Anatoliy; Tanklevskiy, Leonid

2017-06-01

The article deals with innovative technical solutions that provide fire safety in inhabited pressurized compartments of manned spacecraft by means of a fireproof device of inhabited pressurized compartments via application of engineering means of fire prevention and fire spreading prevention by lowering fire load in an inhabited pressurized module up to the point when the maximum possible levels of fire factors in an inhabited pressurized compartment of a manned spacecraft are prevented. Represented technical solutions are used at the present time according to stated recommendations during provision of fire safety of equipment created by a number of Russian organizations for equipage of inhabited pressurized compartments of spacecraft of the Russian segment of International space station.

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

Russian EVA 33

NASA Image and Video Library

2013-06-24

View of Russian cosmonaut Alexander Misurkin (bottom center), Expedition 36 flight engineer, participating in Russian extravehicular activity (EVA) 33. Also visible are the Progress spacecraft docked to the Pirs Docking Compartment (DC1) with the Service Module (SM) .

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.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011479 (24 June 2013) --- Russian cosmonaut Fyodor Yurchikhin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Yurchikhin and Russian cosmonaut Alexander Misurkin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011459 (24 June 2013) --- Russian cosmonaut Fyodor Yurchikhin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Yurchikhin and Russian cosmonaut Alexander Misurkin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011481 (24 June 2013) --- Russian cosmonaut Fyodor Yurchikhin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Yurchikhin and Russian cosmonaut Alexander Misurkin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011441 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011747 (24 June 2013) --- Russian cosmonaut Alexander Misurkin (bottom center), Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011642 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011440 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed one new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011480 (24 June 2013) --- Russian cosmonaut Fyodor Yurchikhin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Yurchikhin and Russian cosmonaut Alexander Misurkin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011745 (24 June 2013) --- Russian cosmonaut Alexander Misurkin (bottom center), Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011598 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed one new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011477 (24 June 2013) --- Russian cosmonaut Fyodor Yurchikhin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Yurchikhin and Russian cosmonaut Alexander Misurkin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011439 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed one new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011640 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011608 (24 June 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, participates in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame), Expedition 36 flight engineer, replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

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

Russian EVA 39.

NASA Image and Video Library

2014-08-18

ISS040E099104 (08/18/2014) --- View of Cosmonaut Oleg Artemyev (blue stripe), Expedition 40 flight engineer outside the International Space Station, taken while performing maintenance work on the Russian segment during the Russian EVA 39 on Aug 18 2014.

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

Russian translations for Cochrane.

PubMed

Yudina, E V; Ziganshina, L E

2015-01-01

Cochrane collaboration has made a huge contribution to the development of evidence-based medicine; Cochrane work is the international gold standard of independent, credible and reliable high-quality information in medicine. Over the past 20 years the Cochrane Collaboration helped transforming decision-making in health and reforming it significantly, saving lives and contributing to longevity [1]. Until recently, Cochrane evidence were available only in English, which represents a significant barrier to their wider use in non-English speaking countries. To provide access to evidence, obtained from Cochrane Reviews, for health professionals and general public (from non-English-speaking countries), bypassing language barriers, Cochrane collaboration in 2014 initiated an international project of translating Plain language summaries of Cochrane Reviews into other languages [2, 3]. Russian translations of Plain language summaries were started in May 2014 by the team from Kazan Federal University (Department of Basic and Clinical Pharmacology; 2014-2015 as an Affiliated Centre in Tatarstan of the Nordic Cochrane Centre, since August 2015 as Cochrane Russia, a Russian branch of Cochrane Nordic, Head - Liliya Eugenevna Ziganshina) on a voluntary basis. To assess the quality of Russian translations of Cochrane Plain Language Summaries (PLS) and their potential impact on the Russian speaking community through user feedback with the overarching aim of furthering the translations project. We conducted the continuous online survey via Google Docs. We invited respondents through the electronic Russian language discussion forum on Essential Medicines (E-lek), links to survey on the Russian Cochrane.org website, invitations to Cochrane contributors registered in Archie from potential Russian-speaking countries. We set up the survey in Russian and English. The respondents were asked to respond to the questionnaire regarding the relevance and potential impact of the Cochrane Russian

Russian EVA 39

NASA Image and Video Library

2014-08-18

ISS040E099874 (08/18/2014) --- Cosmonauts Alexander Skvortsov (red stripe - foreground) and Oleg Artemyev (blue stripe - background), Expedition 40 flight engineers, move to the Russian Service Module for repairs during International Space Station Russian EVA 39 on Aug. 18, 2014.

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.

The German/Russian MIR 1997 Mission: An Overview

NASA Technical Reports Server (NTRS)

1997-01-01

Session TP4 includes short reports concerning: (1) Life Science Experiments During the German-Russian Mir '97 Mission; (2) Orthostatic Intolerance Following Microgravity: A Role for Autonomic Dysfunction; (3) Heart Rate Variability and Skin Blood Flow in Man During Orthostatic Stress in Weightlessness; (4) Effects of Microgravity and Lower Body Negative Pressure on Circulatory Drives from Excercising Calf Muscles; (5) The Mir Station in Its Second Decade: Crew Science Operation During Mir '97; (6) Metabolic WARD (Water, Sodium, Calcium, and Bone Metabolism) and Endocrinological Experiments During the Mir '97 Mission; (7) Long-term Monitoring of the Spine-geometry During the Mir '97 Mission: Introduction of a New Method; and (8) Effects of 20 days of Microgravity (German/Russian Mir '97 Mission) on the Mechanical and Electromyographic Characteristics of Explosive Efforts of the Lower Limbs and of Cycloergometric Exercises of Mild to Sprint-Like Intensity.

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.

Preparation for Testing, Safe Packing and Shipping of Spent Nuclear Fuel from IFIN-HH, Bucharest-Magurele to Russian Federation

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

Dragolici, C.A.; Zorliu, A.; Popa, V.

2007-07-01

The Russian Research Reactor Fuel Return (RRRFR) program is promoted by IAEA and DOE in order to repatriate of irradiated research reactor fuel originally supplied by Russia to facilities outside the country. Developed under the framework of the Global Threat Reduction Initiative (GTRI) the take-back program [1] common goal is to reduce both proliferation and security risks by eliminating or consolidating inventories of high-risk material. The main objective of this program is to support the return to Russian Federation of fresh or irradiated HEU and LEU fuel. Being part of this project, Romania is fulfilling its tasks by examining transportmore » and transfer cask options, assessment of transport routes, and providing cost estimates for required equipment and facility modifications. Spent Nuclear Fuel (SNF) testing, handling, packing and shipping are the most common interests on which the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei' (IFIN-HH) is focusing at the moment. (authors)« less

Expedition 9 Russian News Conference

NASA Image and Video Library

2004-04-20

NASA Deputy Administrator Fred Gregory, far right, and Russian Federal Space Agency Deputy General-Director Nikolai Moiseev, second from right, answer questions from reporters along with other Russian space officials at a news conference, Wednesday, April 21, 2004, at the Russian Mission Control Center outside Moscow following the docking of the Expedition 9 crew and a European Space Agency astronaut to the International Space Station in a Russian Soyuz spacecraft. Photo Credit: (NASA/Bill Ingalls)

Expedition 9 Russian News Conference

NASA Image and Video Library

2004-04-20

NASA Deputy Administrator Fred Gregory, second from right, and Russian Federal Space Agency Deputy General-Director Nikolai Moiseev, center, answer questions from reporters along with other Russian space officials at a news conference, Wednesday, April 21, 2004, at the Russian Mission Control Center outside Moscow following the docking of the Expedition 9 crew and a European Space Agency astronaut to the International Space Station in a Russian Soyuz spacecraft. Photo Credit: (NASA/Bill Ingalls)

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035177 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers. Parts of solar array panels on the orbital outpost are visible in the background,

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035198 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers. A section of the space station is visible in the reflections in his helmet visor.

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035200 (22 Aug. 2013) --- Russian cosmonaut Alexander Misurkin, Expedition 36 flight engineer, attired in a Russian Orlan spacesuit, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 58-minute spacewalk, Misurkin and Russian cosmonaut Fyodor Yurchikhin (out of frame) completed the replacement of a laser communications experiment with a new platform for a small optical camera system, the installation of new spacewalk aids and an inspection of antenna covers. A section of the space station is visible in the reflections in his helmet visor.

Russian EVA-31

NASA Image and Video Library

2012-08-20

ISS032-E-020683 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA fit check

NASA Image and Video Library

2013-06-21

ISS036-E-009793 (21 June 2013) --- Russian cosmonauts Fyodor Yurchikhin (left) and Alexander Misurkin, both Expedition 36 flight engineers, participate in a suited exercise dry run in preparation for a spacewalk in their Russian Orlan spacesuits, which is scheduled for June 24 from the International Space Station’s Pirs docking compartment. Russian cosmonaut Pavel Vinogradov (mostly out of frame at right), Expedition 36 commander, assists Yurchikhin and Misurkin.

Russian Orthography and Learning to Read

ERIC Educational Resources Information Center

Kerek, Eugenia; Niemi, Pekka

2009-01-01

The unique structure of Russian orthography may influence the organization and acquisition of reading skills in Russian. The present review examines phonemic-graphemic correspondences in Russian orthography and discusses its grain-size units and possible difficulties for beginning readers and writers. Russian orthography is governed by a…

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.

Russian Supplementary Dialogues.

ERIC Educational Resources Information Center

Peace Corps, Ashgabat (Turkmenistan).

This manual is designed for the Russian language training of Peace Corps volunteers serving in Turkmenistan, and focuses on daily communication skills needed in that context. It consists of nine topical lessons, each containing several brief dialogues targeting specific language competencies, and exercises. Text is entirely in Russian, except for…

Russian EVA 35

NASA Image and Video Library

2013-08-22

ISS036-E-035256 (22 Aug. 2013) --- Russian cosmonauts Alexander Misurkin (top) and Fyodor Yurchikhin, both Expedition 36 flight engineers, are pictured in the Zvezda Service Module of the International Space Station following a session of extravehicular activity (EVA). Misurkin and Yurchikhin are wearing blue thermal undergarments that complement the Russian Orlan spacesuit.

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

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

NASA Astrophysics Data System (ADS)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

Measurement equivalence of four psychological questionnaires in native-born Germans, Russian-speaking immigrants, and native-born Russians.

PubMed

Hirsch, Oliver; Donner-Banzhoff, Norbert; Bachmann, Viktoria

2013-07-01

Psychological constructs depend on cultural context. It is therefore important to show the equivalence of measurement instruments in cross-cultural research. There is evidence that in Russian-speaking immigrants, cultural and language issues are important in health care. We examined measurement equivalence of the Patient Health Questionnaire-9 (PHQ-9), the Patient Health Questionnaire-15 (PHQ-15), the Hamburg Self-Care Questionnaire (HamSCQ), and the questionnaire on communication preferences of patients with chronic illness (KOPRA) in native-born Germans, Russian-speaking immigrants living in Germany, and native-born Russians living in the former Soviet Union (FSU). All four questionnaires fulfilled requirements of measurement equivalence in confirmatory factor analyses and analyses of differential item functioning. The Russian translations can be used in Russian-speaking immigrants and native-born Russians. This offers further possibilities for cross-cultural research and for an improvement in health care research in Russian-speaking immigrants in Germany. The most pronounced differences occurred in the KOPRA, which point to differences in German and Russian health care systems.

Studying Russian and Soviet History.

ERIC Educational Resources Information Center

Ascher, Abraham, Ed.

These essays were written to assist teachers in the task of making Russian history intelligible to young U.S. students. In "An Approach to Russian History," Edward Keenan proposes that students need to gain a better understanding of how Russians perceive themselves and their history. In "Pre-Petrine Russia," Andrzej S. Kaminski…

English-Russian, Russian-English glossary of coal-cleaning terms

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

Pekar, J.

1987-09-01

The document is an English-Russian, Russian-English glossary of coal-cleaning terms, compiled as a joint U.S./Soviet effort. The need for the glossary resulted from the growing number of language-specific terms used during information exchanges within the framework of the U.S./U.S.S.R. Working Group on Stationary Source Air Pollution Control Technology, under the U.S./U.S.S.R. Agreement of Cooperation in the Field of Environmental Protection.

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

Russian EVA 36.

NASA Image and Video Library

2013-11-09

ISS037-E-028076 (9 Nov. 2013) --- Russian cosmonaut Sergey Ryazanskiy, Expedition 37 flight engineer, attired in a Russian Orlan spacesuit, uses a digital still camera to expose a photo of his helmet visor during a session of extravehicular activity (EVA) as work continues on the International Space Station. Also visible in the reflections in the visor are Russian cosmonaut Oleg Kotov, flight engineer, and various components of the space station and a blue and white portion of Earth. During the five-hour, 50-minute spacewalk, Kotov and Ryazanskiy continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22.

Reading L2 Russian: The Challenges of the Russian-English Dictionary

ERIC Educational Resources Information Center

Comer, William J.

2014-01-01

This descriptive study examines when and how students use Russian-English dictionaries while reading informational texts in Russian and what success they have with word lookup. The study uses introspective verbal protocols (i.e., think-alouds) to follow how readers construct meaning from two texts while reading them for a limited time first…

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

Application of a Systems Engineering Approach to Support Space Reactor Development

NASA Astrophysics Data System (ADS)

Wold, Scott

2005-02-01

In 1992, approximately 25 Russian and 12 U.S. engineers and technicians were involved in the transport, assembly, inspection, and testing of over 90 tons of Russian equipment associated with the Thermionic System Evaluation Test (TSET) Facility. The entire Russian Baikal Test Stand, consisting of a 5.79 m tall vacuum chamber and related support equipment, was reassembled and tested at the TSET facility in less than four months. In November 1992, the first non-nuclear operational test of a complete thermionic power reactor system in the U.S. was accomplished three months ahead of schedule and under budget. A major factor in this accomplishment was the application of a disciplined top-down systems engineering approach and application of a spiral development model to achieve the desired objectives of the TOPAZ International Program (TIP). Systems Engineering is a structured discipline that helps programs and projects conceive, develop, integrate, test and deliver products and services that meet customer requirements within cost and schedule. This paper discusses the impact of Systems Engineering and a spiral development model on the success of the TOPAZ International Program and how the application of a similar approach could help ensure the success of future space reactor development projects.

Comprehensive Experiments on Subcritical Assemblies of Cascade Reactor Systems

NASA Astrophysics Data System (ADS)

Zavyalov, N. V.; Il'kaev, R. I.; Kolesov, V. F.; Ivanin, I. A.; Zhitnik, A. K.; Kuvshinov, M. I.; Nefedov, Yu. Ya.; Punin, V. T.; Tel'nov, A. V.; Khoruzhi, V. Kh.

2017-12-01

Cascade reactors attract particular attention because of their capability of improving the parameters of pulsed reactors and achieving the feasibility of electronuclear facilities. The paper presents the results of three series of experiments on uranium-neptunium cascade assemblies at the Institute of Nuclear and Radiation Physics of the All-Russian Research Institute of Experimental Physics conducted in 2003-2004. The experiments confirmed theoretical conclusions on positive properties of cascade blankets and effectiveness of using neptunium-237 as a means of creating a one-sided connection between the sections.

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.

Russian BAR/EXPERT experiment

NASA Image and Video Library

2009-08-28

ISS020-E-035016 (27 Aug. 2009) --- Russian cosmonaut Gennady Padalka, Expedition 20 commander, uses the Russian BAR/EXPERT science payload to take various environmental measurements in the Zvezda Service Module of the International Space Station.

Russian BAR/EXPERT experiment

NASA Image and Video Library

2009-08-28

ISS020-E-035017 (27 Aug. 2009) --- Russian cosmonaut Gennady Padalka, Expedition 20 commander, uses the Russian BAR/EXPERT science payload to take various environmental measurements in the Zvezda Service Module of the International Space Station.

NATO-Russian Relations: Status and Prospectives

DTIC Science & Technology

2004-12-01

commission to draft it until July 1994. 91 Anne C. Aldis and Roger N. McDermott, Russian Military Reform 1992... Russian Military Reform 1992-2002, p. 4. 42 Finally, the President signed the first National Security Concept of the Russian Federation on 17 December...INTENTIONALLY LEFT BLANK 73 LIST OF REFERENCES Aldis, Anne C. and McDermott, Roger N., Russian Military Reform 1992-2002, (Portland, OR: Conflict

Russian BAR/EXPERT experiment

NASA Image and Video Library

2009-08-28

ISS020-E-035022 (27 Aug. 2009) --- Russian cosmonaut Roman Romanenko, Expedition 20 flight engineer, uses the Russian BAR/EXPERT science payload to take various environmental measurements in the Zvezda Service Module of the International Space Station.

Performance evaluation of the Russian SPT-100 thruster at NASA LeRC

NASA Technical Reports Server (NTRS)

Sankovic, John M.; Hamley, John A.; Haag, Thomas W.

1994-01-01

Performance measurements of a Russian flight-model SPT-100 thruster were obtained as part of a comprehensive program to evaluate engineering issues pertinent to integration with Western spacecraft. Power processing was provided by a US Government developed laboratory power conditioner. When received the thruster had been subjected to only a few hours of acceptance testing by the manufacturer. Accumulated operating time during this study totalled 148 h and included operation of both cathodes. Cathode flow fraction was controlled both manually and using the flow splitter contained within the supplied xenon flow controller. Data were obtained at current levels ranging from 3 A to 5 A and thruster voltages ranging from 200 V to 300 V. Testing centered on the design power of 1.35 kW with a discharge current of 4.5 A. The effects of facility pressure on thruster operation were examined by varying the pressure via injection of xenon into the vacuum chamber. The facility pressure had a significant effect on thruster performance and stability at the conditions tested. Periods of current instabilities were noted throughout the testing period and became more frequent as testing progressed. Performance during periods of stability agreed with previous data obtained in Russian laboratories.

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

Russian expats seek research reforms

NASA Astrophysics Data System (ADS)

Banks, Michael

2009-11-01

Over 170 Russian researchers working abroad have signed a letter addressed to the Russian president, Dmitry Medvedev, and prime minister Vladimir Putin raising concerns about "the catastrophic state of basic science" in the country. The letter, which appeared last month in the Moscow business paper Vedomosti, warns Russian leaders that unless urgent measures are implemented by the government, then science in the country may collapse.

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.

Russian Contract Procurement Document

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

Tobin, J G

2010-03-29

This contract supports the enhancement of physical protection or nuclear material control and accounting systems at institutes or enterprises of the newly independent states under the material protection control and accounting (MPC&A) program. The contract is entered into pursuant to the MPC&A Program, a gratuitous technical assistance program, in accordance with the bilateral Agreements between the Russian Federation and the United States of America concerning the Safe and Secure Transportation, Storage and Destruction of Weapons and the Prevention of Weapons Proliferation of June 1992, as extended and amended by Protocol signed of June 1999, Agreement between the Government of themore » Russian Federation regarding Cooperation in the Area of Nuclear Materials Physical Protection, Control and Accounting of October 1999 and the Russian Federation law of May 1999 on the taxation exemption of gratuitous technical assistance with Russian Federation under registration No.DOE001000.« less

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011590 (24 June 2013) --- Russian cosmonauts Alexander Misurkin (left) and Fyodor Yurchikhin, both Expedition 36 flight engineers, participate in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Yurchikhin replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed a new one.

Russian EVA 33

NASA Image and Video Library

2013-06-24

ISS036-E-011593 (24 June 2013) --- Russian cosmonauts Alexander Misurkin (left) and Fyodor Yurchikhin, both Expedition 36 flight engineers, participate in a session of extravehicular activity (EVA) as work continues on the International Space Station. During the six-hour, 34-minute spacewalk, Misurkin and Yurchikhin replaced an aging fluid flow control panel on the station's Zarya module as preventative maintenance on the cooling system for the Russian segment of the station. They also installed clamps for future power cables as an early step toward swapping the Pirs airlock with a new multipurpose laboratory module. The Russian Federal Space Agency plans to launch a combination research facility, airlock and docking port late this year on a Proton rocket. Yurchikhin and Misurkin also retrieved two science experiments and installed one new one.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020596 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, deploys a small ball-shaped science satellite during a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, also moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021284 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021296 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021028 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020884 (20 Aug. 2012) --- Russian cosmonaut Yuri Malenchenko, Expedition 32 flight engineer, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Malenchenko and Russian cosmonaut Gennady Padalka (out of frame), commander, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021078 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, uses a still camera during a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020610 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021024 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021058 (20 Aug. 2012) --- Russian cosmonaut Yuri Malenchenko, Expedition 32 flight engineer, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Malenchenko and Russian cosmonaut Gennady Padalka (out of frame), commander, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021085 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020619 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, uses a still camera during a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020576 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020594 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021081 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020601 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, deploys a small ball-shaped science satellite during a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, also moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021072 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, uses a still camera during a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020856 (20 Aug. 2012) --- Russian cosmonaut Yuri Malenchenko, Expedition 32 flight engineer, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Malenchenko and Russian cosmonaut Gennady Padalka (out of frame), commander, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021037 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020581 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021293 (20 Aug. 2012) --- Russian cosmonaut Yuri Malenchenko, Expedition 32 flight engineer, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Malenchenko and Russian cosmonaut Gennady Padalka (out of frame), commander, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-020892 (20 Aug. 2012) --- Russian cosmonaut Yuri Malenchenko, Expedition 32 flight engineer, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Malenchenko and Russian cosmonaut Gennady Padalka (out of frame), commander, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021054 (20 Aug. 2012) --- Russian cosmonaut Yuri Malenchenko, Expedition 32 flight engineer, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Malenchenko and Russian cosmonaut Gennady Padalka (out of frame), commander, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021080 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, participates in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021067 (20 Aug. 2012) --- Russian cosmonaut Gennady Padalka, Expedition 32 commander, uses a still camera during a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Russian cosmonaut Yuri Malenchenko (out of frame), flight engineer, moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian Military Reform: Problems and Challenges

DTIC Science & Technology

2010-03-25

RUSSIAN MILITARY REFORM : PROBLEMS AND CHALLENGES BY LIEUTENANT COLONEL ZURAB AGLADZE Georgian Army...USAWC STRATEGIC REASERCH PROJECT RUSSIAN MILITARY REFORM : PROBLEMS AND CHALLENGES by Lieutenant Colonel...noncommissioned officers still continue to be unsolved. Despite some successes, Russian military reform still faces many challenges that will

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.

ALL RUSSIAN INSTITUTE FOR SCIENTIFIC AND TECHNICAL INFORMATION (VINITI) OF THE RUSSIAN ACADEMY OF SCIENCES

PubMed Central

Markusova, Valentina

2012-01-01

The aim of the paper is to overview the leading information processing domain in Russia and Eastern Europe, namely All Russian Institute for Scientific and Technical Information (VINITI ) of the Russian Academy of Sciences. Russian science structure is different from that in the Western Europe and the US. The main aim of VINITI is to collect, process and disseminate scientific information on various fields of science and technology, published in 70 countries in 40 languages, selected from books, journals, conference proceedings, and patents. A special attention is given to the journal selection and depositing manuscripts (a kind of grey literature), an important source for Russian research. VINITI has created the largest database containing about 30 million records dating back to 1980. About 80,000-100,000 new records are added monthly. VINITI publishes the Journal Abstract (JA) on 19 fields of science, including medicine, containing about a million publications annually. Two thirds of these records are foreign and 36.7% – Russian sources. PMID:23322964

All Russian institute for scientific and technical information (viniti) of the Russian academy of sciences.

PubMed

Markusova, Valentina

2012-06-01

The aim of the paper is to overview the leading information processing domain in Russia and Eastern Europe, namely All Russian Institute for Scientific and Technical Information (VINITI ) of the Russian Academy of Sciences. Russian science structure is different from that in the Western Europe and the US. The main aim of VINITI is to collect, process and disseminate scientific information on various fields of science and technology, published in 70 countries in 40 languages, selected from books, journals, conference proceedings, and patents. A special attention is given to the journal selection and depositing manuscripts (a kind of grey literature), an important source for Russian research. VINITI has created the largest database containing about 30 million records dating back to 1980. About 80,000-100,000 new records are added monthly. VINITI publishes the Journal Abstract (JA) on 19 fields of science, including medicine, containing about a million publications annually. Two thirds of these records are foreign and 36.7% - Russian sources.

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.

Russian Space Suits ready

NASA Image and Video Library

2014-08-17

ISS040-E-095609 (17 Aug. 2014) --- Unoccupied Russian Orlan spacesuits for Russian cosmonauts Oleg Artemyev (blue stripes) and Alexander Skvortsov (red stripes), both Expedition 40 flight engineers, are pictured in the Pirs Docking Compartment of the International Space Station on the eve of the spacewalk scheduled for Aug. 18, 2014.

Russian Space Suits ready

NASA Image and Video Library

2014-08-17

ISS040-E-095619 (17 Aug. 2014) --- Unoccupied Russian Orlan spacesuits for Russian cosmonauts Oleg Artemyev (blue stripes) and Alexander Skvortsov (red stripes), both Expedition 40 flight engineers, are pictured in the Pirs Docking Compartment of the International Space Station on the eve of the spacewalk scheduled for Aug. 18, 2014.

Russian Space Suits ready

NASA Image and Video Library

2014-08-17

ISS040-E-095615 (17 Aug. 2014) --- Unoccupied Russian Orlan spacesuits for Russian cosmonauts Oleg Artemyev (blue stripes) and Alexander Skvortsov (red stripes), both Expedition 40 flight engineers, are pictured in the Pirs Docking Compartment of the International Space Station on the eve of the spacewalk scheduled for Aug. 18, 2014.

Russian Space Suits ready

NASA Image and Video Library

2014-08-17

ISS040-E-095617 (17 Aug. 2014) --- Unoccupied Russian Orlan spacesuits for Russian cosmonauts Oleg Artemyev (blue stripes) and Alexander Skvortsov (red stripes), both Expedition 40 flight engineers, are pictured in the Pirs Docking Compartment of the International Space Station on the eve of the spacewalk scheduled for Aug. 18, 2014.

Russian Space Suits ready

NASA Image and Video Library

2014-08-17

ISS040-E-095612 (17 Aug. 2014) --- Unoccupied Russian Orlan spacesuits for Russian cosmonauts Oleg Artemyev (blue stripes) and Alexander Skvortsov (red stripes), both Expedition 40 flight engineers, are pictured in the Pirs Docking Compartment of the International Space Station on the eve of the spacewalk scheduled for Aug. 18, 2014.

Russian electrometallurgy: Achievements, problems, prospects

NASA Astrophysics Data System (ADS)

Utochkin, Yu. I.; Semin, A. E.

2011-12-01

The changes in the Russian metallurgy, in particular, electric furnace steelmaking, having occurred in the recent years are analyzed. The main increase in the steelmaking output is due to putting into operation of new electric furnaces in new miniworks and enterprises equipped earlier with open-hearth furnaces. Reaching the rated capacity of a furnace in Russia substantially lags behind foreign enterprises. Only 30-35% of the Russian market of corrosion-resistant steel are provided by Russian metal.

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.

Russian EVA fit check

NASA Image and Video Library

2013-06-21

ISS036-E-009797 (21 June 2013) --- Russian cosmonauts Fyodor Yurchikhin (left) and Alexander Misurkin, both Expedition 36 flight engineers, participate in a suited exercise dry run in preparation for a spacewalk in their Russian Orlan spacesuits, which is scheduled for June 24 from the International Space Station’s Pirs docking compartment.

Russian EVA no. 39.

NASA Image and Video Library

2014-08-18

ISS040E099355 (08/18/2014) --- Russian cosmonaut Alexander Skvortsov (red stripes), Expedition 40 flight engineer, attired in a Russian Orlan spacesuit outside the International Space Station, participates in a session of extravehicular activity (EVA) number 39 in support of science and maintenance. The Solar array is visible in the background.

Expedition 9 Russian News Conference

NASA Image and Video Library

2004-04-20

NASA Deputy Administrator Fred Gregory, right, and Nikolai Moiseev, Deputy General-Director of the Russian Federal Space Agency, center, share a light-hearted moment at the Russian Mission Control Center outside Moscow, Wednesday, April 21, 2004, following the successful docking of a Russian Soyuz spacecraft to the International Space Station. The Soyuz brought the new Expedition 9 crew and a European Space Agency researcher to the Station following their launch from the Baikonur Cosmodrome in Kazakhstan. Photo Credit: (NASA/Bill Ingalls)

Russian: An Active Introduction.

ERIC Educational Resources Information Center

De La Cruz, Nina

The Active Introduction is one of the modules in an array of materials used in Russian training for beginners at the Foreign Service Institute. It is essentially a catalog of sentences relating to typical daily activities which can be combined to form different communication sequences in dialog form. Students learn to speak Russian through…

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.

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

NASA Astrophysics Data System (ADS)

Shmelev, A. N.; Kulikov, G. G.

2016-12-01

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U-Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction 232+233+234U and 231Pa are formulated. (1) The fuel cycle would shift from fissile 235U to 233U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most "protected" in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of 231Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian

Russian Language Analysis Project

ERIC Educational Resources Information Center

Serianni, Barbara; Rethwisch, Carolyn

2011-01-01

This paper is the result of a language analysis research project focused on the Russian Language. The study included a diverse literature review that included published materials as well as online sources in addition to an interview with a native Russian speaker residing in the United States. Areas of study include the origin and history of the…

Education Standards and Assessment in the Russian Federation: Results from Russian-Dutch Cooperation in Education. First Edition.

ERIC Educational Resources Information Center

Voogt, Joke, Ed.; Plomp, Tjeerd, Ed.

From 1992 to 1997, a joint Russian-Dutch project team contributed to the elaboration of education standards and an educational assessment system in the Russian Federation in the wake of the transition in the Russian Federation from a centralized to a decentralized society. The chapters in this book describe the results of the project and provide…

Proceedings -- US Russian workshop on fuel cell technologies (in English;Russian)

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

Baker, B.; Sylwester, A.

1996-04-01

On September 26--28, 1995, Sandia National Laboratories sponsored the first Joint US/Russian Workshop on Fuel Cell Technology at the Marriott Hotel in Albuquerque, New Mexico. This workshop brought together the US and Russian fuel cell communities as represented by users, producers, R and D establishments and government agencies. Customer needs and potential markets in both countries were discussed to establish a customer focus for the workshop. Parallel technical sessions defined research needs and opportunities for collaboration to advance fuel cell technology. A desired outcome of the workshop was the formation of a Russian/American Fuel Cell Consortium to advance fuel cellmore » technology for application in emerging markets in both countries. This consortium is envisioned to involve industry and national labs in both countries. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.« less

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.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

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

Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR.

PubMed

Mitchell, G E; Furman, W I; Lychagin, E V; Muzichka, A Yu; Nekhaev, G V; Strelkov, A V; Sharapov, E I; Shvetsov, V N; Chernuhin, Yu I; Levakov, B G; Litvin, V I; Lyzhin, A E; Magda, E P; Crawford, B E; Stephenson, S L; Howell, C R; Tornow, W

2005-01-01

Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 10(18)/cm(2)s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286.

Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR

PubMed Central

Mitchell, G. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Strelkov, A. V.; Sharapov, E. I.; Shvetsov, V. N.; Chernuhin, Yu. I.; Levakov, B. G.; Litvin, V. I.; Lyzhin, A. E.; Magda, E. P.; Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Tornow, W

2005-01-01

Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 1018/cm2s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286. PMID:27308126

Russian perspectives: The past shapes the present

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

Houck, R.P.

1994-11-01

This document contains an outline of a speech given to a group of professionals at Pacific Northwest Laboratory which was intended to give an unbiased view of Soviet perceptions. Topics discussed include: The new mission of US and Soviet labs and institutions to develop products and dedicate research to post cold war threat, historical prospectives of Russia, Russian military roles and missions, ideology of Russian politics, evils of capitalism, Russian civil war, communism, world war II, Russian losses during the war, the cold war, reasons why America should care what happens in Russia, the internal threat against a market-based economy,more » the US should help, and the Russian people and their attitudes.« less

A Bright Future for Russian Higher Education.

ERIC Educational Resources Information Center

Kishkovsky, Sophia

2000-01-01

Russian higher education is reinventing itself, but in distinctly Russian ways. While universities embrace new technology, students are frustrated as new freedoms confront old educational methods (grueling entrance exams, rote learning, dictatorial professors, and minimal participation). Education is still basically Russian, though capitalism is…

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

Development of Clinical Pharmacology in the Russian Federation.

PubMed

Petrov, V I; Kagramanyan, I N; Khokhlov, A L; Frolov, M U; Lileeva, E G

2016-05-01

The article aims to provide the history, organization, and approaches to clinical pharmacology in the Russian Federation. This article is based on major international and Russian documents, along with groundbreaking historical facts and scientific articles related to the development of modern clinical pharmacology the Russian Federation. Improving the quality of drug therapy is the main goal of clinical pharmacology in the Russian Federation. Decisions of the World Health Organization, scientific achievements, and the work of well-known scientists among the world community and in the Russian Federation have strongly influenced the development of clinical pharmacology the Russian Federation. Clinical pharmacology in the Russian Federation addresses a wide range of problems; it actively engages in modern scientific research, education; and clinical practice. Clinical pharmacologists participate in studies of new drugs and often have a specific area of expertise. The future development of clinical pharmacology in the Russian Federation will be related to improvements in training, refinement of the framework that regulates clinical pharmacologists, and the creation of clinical pharmacology laboratories with modern equipment. Copyright © 2016 Elsevier HS Journals, Inc. All rights reserved.

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)

Russian EVA 36

NASA Image and Video Library

2013-11-09

ISS037-E-028787 (9 Nov. 2013) --- Russian cosmonauts Oleg Kotov (left) and Sergey Ryazanskiy, both Expedition 37 flight engineers, attired in Russian Orlan spacesuits, participate in a session of extravehicular activity (EVA) in support of assembly and maintenance on the International Space Station. During the five-hour, 50-minute spacewalk, Kotov and Ryazanskiy continued the setup of a combination EVA workstation and biaxial pointing platform that was installed during an Expedition 36 spacewalk on Aug. 22. Earth’s horizon and the blackness of space provide the backdrop for the scene.

Multifractal structures for the Russian stock market

NASA Astrophysics Data System (ADS)

Ikeda, Taro

2018-02-01

In this paper, we apply the multifractal detrended fluctuation analysis (MFDFA) to the Russian stock price returns. To the best of our knowledge, this paper is the first to reveal the multifractal structures for the Russian stock market by financial crises. The contributions of the paper are twofold. (i) Finding the multifractal structures for the Russian stock market. The generalized Hurst exponents estimated become highly-nonlinear to the order of the fluctuation functions. (ii) Computing the multifractality degree according to Zunino et al. (2008). We find that the multifractality degree of the Russian stock market can be categorized within emerging markets, however, the Russian 1998 crisis and the global financial crisis dampen the degree when we consider the order of the polynomial trends in the MFDFA.

An improved Four-Russians method and sparsified Four-Russians algorithm for RNA folding.

PubMed

Frid, Yelena; Gusfield, Dan

2016-01-01

The basic RNA secondary structure prediction problem or single sequence folding problem (SSF) was solved 35 years ago by a now well-known [Formula: see text]-time dynamic programming method. Recently three methodologies-Valiant, Four-Russians, and Sparsification-have been applied to speedup RNA secondary structure prediction. The sparsification method exploits two properties of the input: the number of subsequence Z with the endpoints belonging to the optimal folding set and the maximum number base-pairs L. These sparsity properties satisfy [Formula: see text] and [Formula: see text], and the method reduces the algorithmic running time to O(LZ). While the Four-Russians method utilizes tabling partial results. In this paper, we explore three different algorithmic speedups. We first expand the reformulate the single sequence folding Four-Russians [Formula: see text]-time algorithm, to utilize an on-demand lookup table. Second, we create a framework that combines the fastest Sparsification and new fastest on-demand Four-Russians methods. This combined method has worst-case running time of [Formula: see text], where [Formula: see text] and [Formula: see text]. Third we update the Four-Russians formulation to achieve an on-demand [Formula: see text]-time parallel algorithm. This then leads to an asymptotic speedup of [Formula: see text] where [Formula: see text] and [Formula: see text] the number of subsequence with the endpoint j belonging to the optimal folding set. The on-demand formulation not only removes all extraneous computation and allows us to incorporate more realistic scoring schemes, but leads us to take advantage of the sparsity properties. Through asymptotic analysis and empirical testing on the base-pair maximization variant and a more biologically informative scoring scheme, we show that this Sparse Four-Russians framework is able to achieve a speedup on every problem instance, that is asymptotically never worse, and empirically better than achieved by

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

Russian History; A Guide to Reference Sources.

ERIC Educational Resources Information Center

McGill Univ., Montreal (Quebec). McLennan Library.

This guide identifies reference sources for the study of Russian and Soviet history available in the McGill University (Montreal) McLennan Library. Russian, English, French, and German language works covering Russian history from its origins to World War II are included. The guide is arranged in two parts: general reference sources and…

Update on the direct n-n scattering experiment at the reactor YAGUAR

NASA Astrophysics Data System (ADS)

Stephenson, S. L.; Crawford, B. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Sharapov, E. I.; Shvetsov, V. N.; Strelkov, A. V.; Levakov, B. G.; Lyzhin, A. E.; Chernukhin, Yu. I.; Howell, C. R.; Mitchell, G. E.; Tornow, W.; Showalter-Bucher, R. A.

2013-10-01

The first direct measurement of the 1S0 neutron-neutron scattering experiment using the YAGUAR aperiodic reactor at the Russian Federal Nuclear Center - All Russian Research Institute of Technical Physics has preliminary results. Thermal neutrons are scattered from a thermal neutron ``gas'' within the scattering chamber of the reactor and measured via time-of-flight. These initial results show an unexpectedly large thermal neutron background now understood to be from radiation-induced desorption within the scattering chamber. Analysis of the neutron time-of-flight spectra suggests neutron scattering from H2 and possibly H2O molecules. An experimental value for the desorption yield ηγ of 0.02 molecules/gamma agrees with modeled results. Techniques to reduce the effect of the nonthermal desorption will be presented. This work was supported in part by ISTC project No. 2286, Russia Found. Grant 01-02-17181, the US DOE grants Nos. DE-FG02-97-ER41042 and DE-FG02-97-ER41033, and by the US NSF through Award Nos. 0107263 and 0555652.

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

Russian Higher Education in China

ERIC Educational Resources Information Center

Khisamutdinov, A. A.

2016-01-01

In the first half of the twentieth century, Russian emigrants in China established a network of higher education institutions based on the prerevolutionary Russian educational system. By referring to memoirs and publications in the periodical press, the author traces the history of the most significant educational establishments: Harbin School of…

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 Role of the Russian Methodological and Training Center in providing Nondestructive Assay Technical Assistance to Russian Enterprises

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

Bogdanov, Sergey; Bezhunov, Gennady; Ryazanov, Boris

The Russian Methodological and Training center (RMTC) was initially created to provide not only personnel training in the areas of nuclear material control and accounting (MC&A), but also methodological and technical assistance to the Russian government and nuclear facilities. The goal of the assistance was to promote enhancement of Russian MC&A infrastructure and modernize the MC&A systems at individual enterprises and facilities.

Wiseman and Suraev in Russian segment

NASA Image and Video Library

2014-06-06

ISS040-E-008030 (6 June 2014) --- NASA astronaut Reid Wiseman and Russian cosmonaut Maxim Suraev (background), both Expedition 40 flight engineers, are pictured in the Russian segment of the International Space Station.

Russian: A Guide to the Spoken Language.

ERIC Educational Resources Information Center

Department of Defense, Washington, DC.

This Russian language guide is of assistance in carrying on simple conversations in Russian and is used in conjunction with records. Russian ranks after Chinese and English as the third most widespread language in the world. All the words and phrases are written in a simplified spelling which is read like English. Useful words and phrases include…

Russian Loanword Adaptation in Persian; Optimal Approach

ERIC Educational Resources Information Center

Kambuziya, Aliye Kord Zafaranlu; Hashemi, Eftekhar Sadat

2011-01-01

In this paper we analyzed some of the phonological rules of Russian loanword adaptation in Persian, on the view of Optimal Theory (OT) (Prince & Smolensky, 1993/2004). It is the first study of phonological process on Russian loanwords adaptation in Persian. By gathering about 50 current Russian loanwords, we selected some of them to analyze. We…

Genetic stock identification of Russian honey bees.

PubMed

Bourgeois, Lelania; Sheppard, Walter S; Sylvester, H Allen; Rinderer, Thomas E

2010-06-01

A genetic stock certification assay was developed to distinguish Russian honey bees from other European (Apis mellifera L.) stocks that are commercially produced in the United States. In total, 11 microsatellite and five single-nucleotide polymorphism loci were used. Loci were selected for relatively high levels of homogeneity within each group and for differences in allele frequencies between groups. A baseline sample consisted of the 18 lines of Russian honey bees released to the Russian Bee Breeders Association and bees from 34 queen breeders representing commercially produced European honey bee stocks. Suitability tests of the baseline sample pool showed high levels of accuracy. The probability of correct assignment was 94.2% for non-Russian bees and 93.3% for Russian bees. A neighbor-joining phenogram representing genetic distance data showed clear distinction of Russian and non-Russian honey bee stocks. Furthermore, a test of appropriate sample size showed a sample of eight bees per colony maximizes accuracy and consistency of the results. An additional 34 samples were tested as blind samples (origin unknown to those collecting data) to determine accuracy of individual assignment tests. Only one of these samples was incorrectly assigned. The 18 current breeding lines were represented among the 2009 blind sampling, demonstrating temporal stability of the genetic stock identification assay. The certification assay will be used through services provided by a service laboratory, by the Russian Bee Breeders Association to genetically certify their stock. The genetic certification will be used in conjunction with continued selection for favorable traits, such as honey production and varroa and tracheal mite resistance.

[Bioethics in Russian neurology and epileptology].

PubMed

Mikhalkovska-Karlova, E P

2016-01-01

Historical roots and further development of bioethics in domestic neurology and epileptology are considered. The main bioethical principles were established during the formation of the Russian clinical school and neurosciences. It is most distinctly seen in the development of bioethics in neurology and epileptology. In the author's opinion, the Russian scientist V.M. Bekhterev had played a prominent role in the field. In the time when the term "bioethics" was not coined and its principles were not formulated, V.M. Bekhterev had created the Russian league against epilepsy and established the foundations of the International League Against Epilepsy (ILAE) as the organizations working on the problems of medical and social care to patients with epilepsy. In Russia, the Russian society of neurologists has been doing a great work in the field.

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.

31 CFR 540.304 - Government of the Russian Federation.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Government of the Russian Federation...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.304 Government of the Russian Federation. (a) The term Government of the Russian Federation means the Government of the Russian Federation, any...

31 CFR 540.304 - Government of the Russian Federation.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Government of the Russian Federation...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.304 Government of the Russian Federation. (a) The term Government of the Russian Federation means the Government of the Russian Federation, any...

31 CFR 540.304 - Government of the Russian Federation.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Government of the Russian Federation...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.304 Government of the Russian Federation. (a) The term Government of the Russian Federation means the Government of the Russian Federation, any...

31 CFR 540.304 - Government of the Russian Federation.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Government of the Russian Federation...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.304 Government of the Russian Federation. (a) The term Government of the Russian Federation means the Government of the Russian Federation, any...

31 CFR 540.304 - Government of the Russian Federation.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Government of the Russian Federation...) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.304 Government of the Russian Federation. (a) The term Government of the Russian Federation means the Government of the Russian Federation, any...

Reading Russian-English Homographs in Sentence Contexts: Evidence from ERPs

ERIC Educational Resources Information Center

Jouravlev, Olessia; Jared, Debra

2014-01-01

The current study investigated whether Russian--English bilinguals activate knowledge of Russian when reading English sentences. Russian and English share only a few letters, but there are some interlingual homographs (e.g., POT, which means "mouth" in Russian). Critical sentences were written such that the Russian meaning of the…

English/Russian and Russian/English glossary of physical protection terms

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

Soo Hoo, M.S.

1995-07-01

This glossary was prepared in fulfillment of the Glossary Preparation Task identified in the Program Plan for providing Assistance to the Russian Federation in Nuclear Material Control and Accounting and Physical Protection. The Program Plan is part of the Cooperative Threat Reduction Program as provided for under House Resolution (H.R.) 3807 (Title II, as referenced under Public Law (P.L.) 102-229. The terms in this glossary were derived from physical protection training material prepared at Sandia. The training material, and thus refinements to the glossary, has undergone years of development in presentation to both domestic and international audiences. Also, Russian Colleaguesmore » and interpreters have reviewed the translations for accuracy.« less

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.

PLATO Sitcom Dialogs for Russian.

ERIC Educational Resources Information Center

Dawson, Clayton; Provenzano, Nolen

1981-01-01

Situation comedy (sitcom) dialogs that are included in PLATO lessons for first year Russian students are described. These comprehension exercises make use of both the touch panel and the audio capabilities of PLATO. The sitcom dialogs were written by a native speaker of Russian and are based on the vocabulary in the textbook plus a small number of…

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.

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

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

Shmelev, A. N.; Kulikov, G. G., E-mail: ggkulikov@mephi.ru

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results aremore » analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction {sup 232+233+234}U and {sup 231}Pa are formulated. (1) The fuel cycle would shift from fissile {sup 235}U to {sup 233}U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of {sup 231}Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the

Water-quality data for the Russian River Basin, Mendocino and Sonoma Counties, California, 2005-2010

USGS Publications Warehouse

Anders, Robert; Davidek, Karl; Stoeckel, Donald M.

2011-01-01

Field measurements included discharge, barometric pressure, dissolved oxygen, pH, specific conductance, temperature, and turbidity. All samples were analyzed for nutrients, major ions, trace metals, total and dissolved organic carbon, organic wastewater compounds, standard bacterial indicators, and the stable isotopes of hydrogen and oxygen. Standard bacterial indicators included total coliform, Escherichia coli, enterococci, and Clostridium perfringens for the period 2005 through 2007, and total and fecal coliform, and enterococci for 2010. In addition, enrichment of enterococci was performed on all surface-water samples collected during summer 2006, for detection of the human-associated enterococcal surface protein in Enterococcus faecium to assess the presence of sewage effluent in the Russian River. Other analyses included organic wastewater compounds of bed sediment samples collected from four Russian River sites during 2005; carbon-13 isotopic values of the dissolved inorganic carbon for surface-water and groundwater samples collected during 2006; human-use pharmaceuticals on Russian River samples collected during 2007 and 2010; and the radiogenic isotopes tritium and carbon-14 for groundwater samples collected during 2008.

Coverage of Russian psychological contributions in American psychology textbooks.

PubMed

Aleksandrova-Howell, Maria; Abramson, Charles I; Craig, David Philip Arthur

2012-01-01

Internationalizing psychology is an important component of current globalization trends. American textbooks on the history of psychology and introductory psychology were surveyed for the presence of historical and contemporary important Russian psychologists to assess the current status of Russian-American crossfertilization. Of a list of 97 important Russian psychologists, as determined by the editors of the Russian journal Methodology and History in Psychology, less than 22% are mentioned in the reviewed texts. The most common names were Pavlov, Luria, and Vygotsky. As the internet is arguably the single most important factor affecting the increase of international communication and dissemination of knowledge, we also searched for these 97 names on various websites, most notably Wikipedia and Google. Forty-one internet sites contained some amount of biographical information about Russian psychologists. On Wikipedia, 14 Russian psychologists had articles documenting biographical information. We also developed a rubric to determine the amount of information available on the internet for these psychologists and compared Wikipedia's mean score with various other websites. Wikipedia pages on average had a significantly higher score than the rest of the internet. Recommendations to improve Russian coverage in America are provided and include: (1) developing pages on Wikipedia and other virtual venues highlighting Russian contributions, (2) soliciting articles for US journals from Russian psychologists, and (3) incorporating Russian contributions in introductory and historical textbooks. We provide a partial bibliography of Russian contributions that can be used by authors of such textbooks. We would like to thank Dr Viktor Fedorovich Petrenko and Dr Igor Nikolaevich Karitsky from the journal Methodology and History of Psychology for supplying the names of the Russian psychologists. We would also like to express our appreciation to Robert García for reviewing and

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 Russian Way of Warfare

DTIC Science & Technology

2017-01-01

against any of the country’s neighbors, and Russia has invested heavily in intelligence , surveillance, and reconnaissance (ISR) capabilities to...annexation of Crimea and the destabilization of eastern Ukraine in the vicinity of Donetsk and Luhansk. Russian use of Spetsnaz, intelligence services...greater escalation risk. Russian use of Spetsnaz, intelligence services, and paramilitaries could be an important element of a conflict in

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.

Health Care in the Russian Federation.

PubMed

Younger, David S

2016-11-01

The Russian Federation health system has its roots in the country's complex political history. The Ministry of Health and Social Development and its associated federal services are the principal Russian institutions subserving the Russian Federation. Funding for the health system goes through 2 channels: the general revenue budget managed by federal, regional, and local health authorities, and the Mandatory Health Insurance Fund. Although the Soviet Union was the first country in the world to guarantee free medical care as a constitutional right to all its citizens, quality and accessibility are in question. Copyright © 2016 Elsevier Inc. All rights reserved.

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

Russian policy and the Korean crisis. Final report

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

Blank, S.J.

1994-09-30

The author relates Russian policy to the broader struggle for power and coherence in Russian security policy generally, and in Russian policy towards Asia. He also explores the ambivalence of Russia's position, trying to maintain a foot in all camps on the Korean issue even as it asserts its opposition to nuclear proliferation and to any settlement from which it is excluded. The author concludes by expanding his frame of reference to use the Russian case as an example illustrating the difficulties inherent in moving to a new, broader Northeast Asian security system.

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.

Detecting Bots on Russian Political Twitter.

PubMed

Stukal, Denis; Sanovich, Sergey; Bonneau, Richard; Tucker, Joshua A

2017-12-01

Automated and semiautomated Twitter accounts, bots, have recently gained significant public attention due to their potential interference in the political realm. In this study, we develop a methodology for detecting bots on Twitter using an ensemble of classifiers and apply it to study bot activity within political discussions in the Russian Twittersphere. We focus on the interval from February 2014 to December 2015, an especially consequential period in Russian politics. Among accounts actively Tweeting about Russian politics, we find that on the majority of days, the proportion of Tweets produced by bots exceeds 50%. We reveal bot characteristics that distinguish them from humans in this corpus, and find that the software platform used for Tweeting is among the best predictors of bots. Finally, we find suggestive evidence that one prominent activity that bots were involved in on Russian political Twitter is the spread of news stories and promotion of media who produce them.

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.

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.

Predicate Adjective Usage in Standard Russian.

ERIC Educational Resources Information Center

Benson, Morton

1959-01-01

This paper describes predicate adjective usage in modern standard Russian using a corpus of written Russian derived from "Pravda" (neutral literary style) and "Krokodil" (conversational material). The short, long nominative, and instrumental forms are examined in relation to the type of adjective, copulative verb, sentence subject, and other…

Ranking of Russian Higher Education Institutions

ERIC Educational Resources Information Center

Pokholkov, Yuri P.; Chuchalin, Alexander I.; Agranovich, Boris L.; Mogilnitsky, Sergey B.

2007-01-01

This article considers some patterns of ranking higher education institutions which are used in the Russian Federation to reveal strengths and weaknesses in meeting the national individual, societal and state-related needs, as well as those of the international academic community concerning relevant information on Russian higher education…

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.

Selected translated abstracts of Russian-language climate-change publications. 4: General circulation models (in English;Russian)

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

Burtis, M.D.; Razuvaev, V.N.; Sivachok, S.G.

1996-10-01

This report presents English-translated abstracts of important Russian-language literature concerning general circulation models as they relate to climate change. Into addition to the bibliographic citations and abstracts translated into English, this report presents the original citations and abstracts in Russian. Author and title indexes are included to assist the reader in locating abstracts of particular interest.

Selected translated abstracts of Russian-language climate-change publications: II, Clouds. Issue 159 (in English;Russian)

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

Burtis, M.D.

This report presents abstracts (translated into English) of important Russian-language literature concerning clouds as they relate to climate change. In addition to the bibliographic citations and abstracts translated into English, this report presents the original citations and abstracts in Russian. Author and title indexes are included to assist the reader in locating abstracts of particular interest.

Aspects of the Teaching of Russian.

ERIC Educational Resources Information Center

Baker, Robert L.

The process of learning Russian should be no more difficult than the process of learning other languages although it may take somewhat longer. The phonetic system should not present major difficulties with respect to individual sounds, but intonation may be difficult because Russian pitch patterns represent different intentions and emotions than…

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.

Russian Rocket Engine Test

NASA Technical Reports Server (NTRS)

1998-01-01

NASA engineers successfully tested a Russian-built rocket engine on November 4, 1998 at the Marshall Space Flight Center (MSFC) Advanced Engine Test Facility, which had been used for testing the Saturn V F-1 engines and Space Shuttle Main engines. The MSFC was under a Space Act Agreement with Lockheed Martin Astronautics of Denver to provide a series of test firings of the Atlas III propulsion system configured with the Russian-designed RD-180 engine. The tests were designed to measure the performance of the Atlas III propulsion system, which included avionics and propellant tanks and lines, and how these components interacted with the RD-180 engine. The RD-180 is powered by kerosene and liquid oxygen, the same fuel mix used in Saturn rockets. The RD-180, the most powerful rocket engine tested at the MSFC since Saturn rocket tests in the 1960s, generated 860,000 pounds of thrust. The test was the first test ever anywhere outside Russia of a Russian designed and built engine.

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.

Radioactive waste disposal in seas adjacent to the territory of the Russian Federation.

PubMed

Yablokov, A V

2001-01-01

The former USSR illegally dumped into the ocean liquid and solid radioactive wastes (RW) originating from nuclear-powered vessels and ships. The Russian President created a special Commission to analyse both the scale and consequences of this activity. According to documentary data and expert estimates at the Commission's disposal, the maximum activity of RW that entered the seas adjacent to Russian territory could have been as much as 2,500 kCi at the time of disposal. The greatest radio-ecological hazard comes from reactors from nuclear submarines and core plates of the nuclear icebreaker 'Lenin', which had spent nuclear fuel in place and which were dumped in shallow water in the Kara Sea near Novaya Zemlya. Editor's note: This article extracts material from a Commission which published a report produced in Russia in 1993. Numerous sources in many Ministries and other government agencies, noted in the text, formed the basis for the final draft. The authors of the draft report were A. Yablokov, V. Karasev, V. Rumyantsev, M. Kokeev, O. Petrov, V. Lystsov, A. Yemelyanenkov and P. Rubtsov. After approving the draft report, the Commission submitted the report to the President of the Russian Federation in February 1993. By Presidential decision, this report (after several technical corrections) was open to the public: it is known variously as 'the Yablokov Commission report, or more simply the 'Yablokov Report', the 'White Book' or 'Yablokov White Paper'. During April-May 1993, 500 copies were distributed among governmental agencies inside Russia, and abroad through a net of Russian Embassies. This article was later sent to Dr Mike Champ as part of the ongoing collections of papers on the Arctic published in this journal (edited by Champ et al.: 1997 'Contaminants in the Arctic', Marine Pollution Bulletin 35, pp. 203-385 and in Marine Pollution Bulletin 2000, vol. 40, pp. 801-868, and continued with the present collection).

(In)Complete Acquisition of Aspect in Second Language and Heritage Russian

ERIC Educational Resources Information Center

Mikhaylova, Anna

2012-01-01

This dissertation compares the knowledge of Russian Verbal Aspect in two types of learners enrolled in college level Russian courses: foreign language learners of Russian whose native language is English and heritage language speakers of Russian whose dominant language at the time of study is English. Russian Aspect is known to be problematic both…

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).

Roots of Russian Irregular Warfare

DTIC Science & Technology

2016-12-01

ADDRESS(ES) N /A 10. SPONSORING / MONITORING AGENCY REPORT NUMBER 11. SUPPLEMENTARY NOTES The views expressed in this thesis are those of the author...Annual Address to the Federal Assembly of the Russian Federation,” 2005. 99 J. N . Westwood, Endurance and Endeavour: Russian History, 1812–1992 (New...Vladimir N . Brovkin, Behind the Front Lines of the Civil War: Political Parties and Social Movements in Russia, 1918–1922 (Princeton, NJ: Princeton

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.

Violence on the Russian & American Media Screen and Youth Audience

ERIC Educational Resources Information Center

Fedorov, Alexander

2003-01-01

The comparison of the Russian and American experience regarding media violence, standards for rating Russian media programs, and a course of study on media violence for students will have a significant impact upon Russian society, will raise Russian societal and governmental attention to the infringement of the Rights of the Child on the Russian…

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.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021060 (20 Aug. 2012) --- Russian cosmonauts Gennady Padalka (top), Expedition 32 commander; and Yuri Malenchenko, flight engineer, participate in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Malenchenko moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021061 (20 Aug. 2012) --- Russian cosmonauts Gennady Padalka (top), Expedition 32 commander; and Yuri Malenchenko, flight engineer, participate in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Malenchenko moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021044 (20 Aug. 2012) --- Russian cosmonauts Gennady Padalka (top), Expedition 32 commander; and Yuri Malenchenko, flight engineer, participate in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Malenchenko moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021046 (20 Aug. 2012) --- Russian cosmonauts Gennady Padalka (top), Expedition 32 commander; and Yuri Malenchenko, flight engineer, participate in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Malenchenko moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

Russian EVA-31 spacewalk

NASA Image and Video Library

2012-08-20

ISS032-E-021286 (20 Aug. 2012) --- Russian cosmonauts Gennady Padalka (top), Expedition 32 commander; and Yuri Malenchenko, flight engineer, participate in a session of extravehicular activity (EVA) to continue outfitting the International Space Station. During the five-hour, 51-minute spacewalk, Padalka and Malenchenko moved the Strela-2 cargo boom from the Pirs docking compartment to the Zarya module to prepare Pirs for its eventual replacement with a new Russian multipurpose laboratory module. The two spacewalking cosmonauts also installed micrometeoroid debris shields on the exterior of the Zvezda service module and deployed a small science satellite.

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

Impact of Alleged Russian Cyber Attacks

DTIC Science & Technology

2009-05-01

security. 15. SUBJECT TERMS Cyber Security, Cyber Warfare , Estonia, Georgia, Russian Federation Cyber Strategy, Convention on Cybercrime, NATO Center...Federation ......................................................................................... 33  X.  The Future of Russian Cyber Warfare ................................................................... 39...Issue 15.09); Binoy Kampmark, Cyber Warfare Between Estonia And Russia, (Contemporary Review: Autumn, 2003), p 288-293; Jaak Aaviksoo, Address by the

Syllabus for Use in Russian Literature.

ERIC Educational Resources Information Center

Cernonok, Jevgenij

This syllabus outlines a two semester course to accompany the basic textbook: THE EPIC OF RUSSIAN LITERATURE by Marc Slonim. An introduction to the guide gives a brief summary of the history of Russian literature and objectives of the course are stated, defining concepts and understandings to be developed. In addition, teaching techniques are…

Meta-analysis: Problems with Russian Publications.

PubMed

Verbitskaya, E V

2015-01-01

Meta-analysis is a powerful tool to identify Evidence Based medical technologies (interventions) for use in every day practice. Meta-analysis uses statistical approaches to combine results from multiple studies in an effort to increase power (over individual studies), improve estimates of the size of the effect and/or to resolve uncertainty when reports disagree. Meta-analysis is a quantitative, formal study design used to systematically assess previous research studies to derive conclusions from this research. Meta-analysis may provide more precise estimate of the effect of treatment or risk factor for a disease, or other outcomes, than any individual study contributing to the pooled analysis.We have quite a substantial number of Russian medical publications, but not so many Meta-Analyses published in Russian. Russian publications are cited in English language papers not so often. A total of 90% of clinical studies included in published Meta-Analyses incorporate only English language papers. International studies or papers with Russian co-authors are published in English language. The main question is: what is the problem with inclusion of Russian medical publications in Meta-Analysis? The main reasons for this are the following: 1) It is difficult to find Russian papers, difficult to work with them and to work with Russian journals:a. There are single Russian Biomedical Journals, which are translated into English and are included in databases (PubMed, Scopus and other), despite the fact that all of them have English language abstracts.b. The majority the meta-analyses authors use in their work different citation management software such as the Mendeley, Reference Manager, ProCite, EndNote, and others. These citation management systems allow scientists to organize their own literature databases with internet searches and have adds-on for the Office programs what makes process of literature citation very convenient. The Internet sites of the majority of

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.

Saltcedar and Russian Olive Control Demonstration Act Science Assessment

USGS Publications Warehouse

Shafroth, Patrick B.; Brown, Curtis A.; Merritt, David M.

2010-01-01

The primary intent of this document is to provide the science assessment called for under The Saltcedar and Russian Olive Control Demonstration Act of 2006 (Public Law 109-320; the Act). A secondary purpose is to provide a common background for applicants for prospective demonstration projects, should funds be appropriated for this second phase of the Act. This document synthesizes the state-of-the-science on the following topics: the distribution and abundance (extent) of saltcedar (Tamarix spp.) and Russian olive (Elaeagnus angustifolia) in the Western United States, potential for water savings associated with controlling saltcedar and Russian olive and the associated restoration of occupied sites, considerations related to wildlife use of saltcedar and Russian olive habitat or restored habitats, methods to control saltcedar and Russian olive, possible utilization of dead biomass following removal of saltcedar and Russian olive, and approaches and challenges associated with revegetation or restoration following control efforts. A concluding chapter discusses possible long-term management strategies, needs for additional study, potentially useful field demonstration projects, and a planning process for on-the-ground projects involving removal of saltcedar and Russian olive.

Legal Portion in Russian Inheritance Law

ERIC Educational Resources Information Center

Inshina, Roza; Murzalimova, Lyudmila

2013-01-01

In this paper the authors describe the right to inherit as one of the basic human rights guaranteed by the Constitution of the Russian Federation. The state has set rules according to which after a person's death, his or her property is inherited by other persons. The Russian civil legislation establishes the institution of legal portions that is…

Russian Basic Course: Reader, Volume II.

ERIC Educational Resources Information Center

Defense Language Inst., Washington, DC.

This anthology of short stories is representative of well-known 19th and 20th century Russian writers. Eleven stories, often adapted or abridged, are arranged in order of increasing difficulty and intended for use in intermediate and advanced phases of the Russian Basic Course. The selections, all in Cyrillic script, include: (1) A. S. Pushkin's…

Common fungal diseases of Russian forests

Treesearch

Evgeny P. Kuz' michevl; Ella s. Sokolova; Elena G. Kulikova

2001-01-01

Describes common fungal diseases of Russian forests, including diagnostic signs and symptoms, pathogen biology, damage caused by the disease, and methods of control. The fungal diseases are divided into two groups: those that are the most common in Russian forests and those that are found only in Russia. Within each group, diseases are subdivided by plant organ...

Conversational Russian in Russkij Jazyk Za Rubezom

ERIC Educational Resources Information Center

Morgan, Thomas W.

1976-01-01

Reviews materials relating to Russian speech which have appeared in the journal "Russkij Jazyk Za Rubezom" since its first issue in 1967. The articles reviewed are divided into three categories: the stylistics of Russian speech, specific conversation expressions, and methodology. (CLK)

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.

An investigation of tritium transfer in reactor loops

NASA Astrophysics Data System (ADS)

Ilyasova, O. H.; Mosunova, N. A.

2017-09-01

The work is devoted to the important task of the numerical simulation and analysis of the tritium behaviour in the reactor loops. The simulation was carried out by HYDRA-IBRAE/LM code, which is being developed in Nuclear safety institute of the Russian Academy of Sciences. The code is intended for modeling of the liquid metal flow (sodium, lead and lead-bismuth) on the base of non-homogeneous and non-equilibrium two-fluid model. In order to simulate tritium transfer in the code, the special module has been developed. Module includes the models describing the main phenomena of tritium behaviour in reactor loops: transfer, permeation, leakage, etc. Because of shortage of the experimental data, a lot of analytical tests and comparative calculations were considered. Some of them are presented in this work. The comparison of estimation results and experimental and analytical data demonstrate not only qualitative but also good quantitative agreement. It is possible to confirm that HYDRA-IBRAE/LM code allows modeling tritium transfer in reactor loops.

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.

Russian Orthodox Culture or Russian Orthodox Teaching? Reflections on the Textbooks in Religious Education in Contemporary Russia

ERIC Educational Resources Information Center

Shnirelman, Victor A.

2012-01-01

A new textbook in religious education is analysed, first, in comparison to earlier ones, and, second, with respect to its content and goals. The aim of the Russian Orthodox Church (ROC) is threefold--to educate into religion, to foster the ethnic Russian identity and to make youngsters loyal to the state. Actually, the state rather than society is…

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.

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.

Russian Media Education Researches: 1950-2010

ERIC Educational Resources Information Center

Federov, Alexander

2010-01-01

This article analyzed the development of Russian media education researches from 1950 to 2010 years. The list of theses of the Russian authors on the subject of Media Education is about 180 titles since 1950. Nearly 70 of them have been defended for the recent 10 years. From 1950 till 1959 six theses were defended, from 1960 till 1969--15; from…

Ships in Russian Literature: Folklore Aesthetics

ERIC Educational Resources Information Center

Dudareva, Marianna A.; Pogukaeva, Anna V.; Polyantseva, Evgeniya A.; Karpova, Yulia V.

2017-01-01

The paper studies a genesis of the ship image in the Russian literature and folklore, an idea of "other kingdom" in the Russian literature poetics of the 19-20 centuries. An emphasis is put on the issues related to the metaphor of a ship, a boat in the artistic world of Lermontov, Turgenev, Dostoevsky and in the poetry of the early 20th…

Commonalities in Russian Military Operations in Urban Environments

DTIC Science & Technology

2003-06-06

6th Army was the Russian 62nd Siberian Army commanded by Vasili I Chuikov. The 6th Siberian Army was demoralized and depleted after a year of warfare...Battle of Stalingrad pitted untrained Russian militia and civilians against a highly trained German Army. Predictably the Russians experienced...Pennsylvania: U.S. Army War College. Chuikov, Vasili I. 1964. The Battle for Stalingrad, Trans. by: Harold Silver, New York, New York: Holt, Rhinehart and

Assessing Russian Reactions to U.S. Missile Defense

DTIC Science & Technology

2001-09-01

International Studies, 2001)18, 27. 108 See ibid., 14; and Alexander Savelyev , “Russian Strategic Forces: Their Future and the Issue of BMD,” A...International Peace; Monterey, CA: Monterey Institute of International Studies, 2001) 24, 26, 28. 110 Alexander Savelyev , “Russian Strategic Forces: Their...example, Alexander Savelyev of the Russian Academy of Sciences has made the point that “by all the standards of ‘strategic stability’ the deployment of

The consequences of political dictatorship for Russian science.

PubMed

Soyfer, V N

2001-09-01

The Soviet communist regime had devastating consequences on the state of Russian twentieth century science. Country Communist leaders promoted Trofim Lysenko--an agronomist and keen supporter of the inheritance of acquired characters--and the Soviet government imposed a complete ban on the practice and teaching of genetics, which it condemned as a "bourgeois perversion". Russian science, which had previously flourished, rapidly declined, and many valuable scientific discoveries made by leading Russian geneticists were forgotten.

Russian Emotion Vocabulary in American Learners' Narratives

ERIC Educational Resources Information Center

Pavlenko, Aneta; Driagina, Viktoria

2007-01-01

This study compared the uses of emotion vocabulary in narratives elicited from monolingual speakers of Russian and English and advanced American learners of Russian. Monolingual speakers differed significantly in the distribution of emotion terms across morphosyntactic categories: English speakers favored an adjectival pattern of emotion…

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.

Russian Electoral Politics and the Search for National Identity.

ERIC Educational Resources Information Center

Ishiyama, John T.; Launer, Michael K.; Likhachova, Irina E.; Williams, David Cratis; Young, Marilyn J.

1997-01-01

Explores the role of national identity formation in the democratization of the Russian Federation, analyzing arguments of the 1993 and 1995 Duma elections and the 1996 Russian presidential campaign. Contends results of these elections say more about the search for Russian identity in the wake of social and economic change than they do about the…

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

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.

Social Factors Influencing Russian Male Alcohol Use over the Life Course: A Qualitative Study Investigating Age Based Social Norms, Masculinity, and Workplace Context.

PubMed

Keenan, Katherine; Saburova, Lyudmila; Bobrova, Natalia; Elbourne, Diana; Ashwin, Sarah; Leon, David A

2015-01-01

The massive fluctuations occurring in Russian alcohol-related mortality since the mid-1980s cannot be seen outside of the context of great social and economic change. There is a dearth of qualitative studies about Russian male drinking and especially needed are those that address social processes and individual changes in drinking. Conducted as part of a longitudinal study on men's alcohol consumption in Izhevsk, this qualitative study uses 25 semi-structured biographical interviews with men aged 33-60 years to explore life course variation in drinking. The dominant pattern was decreasing binge and frequent drinking as men reached middle age which was precipitated by family building, reductions in drinking with work colleagues, and health concerns. A minority of men described chaotic drinking histories with periods of abstinence and heavy drinking. The results highlight the importance of the blue-collar work environment for conditioning male heavy drinking in young adulthood through a variety of social, normative and structural mechanisms. Post-Soviet changes had a structural influence on the propensity for workplace drinking but the important social function of male drinking sessions remained. Bonding with workmates through heavy drinking was seen as an unavoidable and essential part of young men's social life. With age peer pressure to drink decreased and the need to perform the role of responsible breadwinner put different behavioural demands on men. For some resisting social pressure to drink became an important site of self-determination and a mark of masculine maturity. Over the lifetime the place where masculine identity was asserted shifted from the workplace to the home, which commonly resulted in a reduction in drinking. We contribute to existing theories of Russian male drinking by showing that the performance of age-related social roles influences Russian men's drinking patterns, drinking contexts and their attitudes. Further research should be

View of EV Crewmember during Russian EVA 29

NASA Image and Video Library

2011-08-03

ISS028-E-020969 (3 Aug. 2011) --- Russian cosmonauts Sergei Volkov and Alexander Samokutyaev (out of frame), both Expedition 28 flight engineers, attired in Russian Orlan spacesuits, participate in a session of extravehicular activity (EVA) on the Russian segment of the International Space Station. During the six-hour, 23-minute spacewalk, Volkov and Samokutyaev moved a cargo boom from one airlock to another, installed a prototype laser communications system and deployed an amateur radio micro-satellite.

Regulation of the Debt Sustainability of the Russian Economy

ERIC Educational Resources Information Center

Seleznev, Alexander Z.; Chapluk, Vladimir Z.; Sayrenko, Tatiana N.; Sorokina, Larisa N.; Pertovskaya, Maria V.; Alekseenko, Elena A.

2016-01-01

The relevance of the investigating problem is caused by the need to reduce the total aggregated amount of debt in Russian economy in conditions of crisis and the strengthening of external anti-Russian sanctions. In this context, the purpose of this article is to identify measures aimed to regulate debt sustainability of the Russian economy using…

Vladimir Zhirinovskiy: An Assessment of a Russian Ultra-Nationalist

DTIC Science & Technology

1994-04-01

establish his party, Zhirinovskiy has stated: "Never. All that is propaganda. Just like they say that I am a Zionist, anti-Jewish, a member of the radical ... right or left, or Jewish. Nothing of the sort. I am Russian, Russian, Russian.ś 3 In response to questions about KGB funding, he has said: "With

Russian Energy Policy Toward Neighboring Countries

DTIC Science & Technology

2010-03-22

countries, Russian energy firms are able to manipulate the internal political situation by favoring certain local businessmen and politicians...Russia is even more significant. It is the single largest contributor to the Russian government’s budget, providing about 25% of tax receipts. It...policy changed in 2003, when Mikhail Khodorkovsky, head of the giant Yukos oil firm, was arrested on tax evasion and other charges. However, most

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

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

Russian Literature: A Guide to Reference Sources.

ERIC Educational Resources Information Center

McGill Univ., Montreal (Quebec). McLennan Library.

An annotated bibliography of general and specialized reference works for Russian and Soviet literature is intended for both students and researchers. English language and Russian language sources in the McGill University (Canada) libraries are included. Subject headings include guides (to the literature and to archival resources), encyclopedias,…

Temperature and pressure dependences of kimberlite melts viscosity (experimental-theoretical study)

NASA Astrophysics Data System (ADS)

Persikov, Eduard; Bykhtiyarov, Pavel; Cokol, Alexsander

2016-04-01

Experimental data on temperature and pressure dependences of viscosity of model kimberlite melts (silicate 82 + carbonate 18, wt. %, 100NBO/T = 313) have been obtained for the first time at 100 MPa of CO2 pressure and at the lithostatic pressures up to 7.5 GPa in the temperature range 1350 oC - 1950 oC using radiation high gas pressure apparatus and press free split-sphere multi - anvil apparatus (BARS). Experimental data obtained on temperature and pressure dependences of viscosity of model kimberlite melts at moderate and high pressures is compared with predicted data on these dependences of viscosity of basaltic melts (100NBO/T = 58) in the same T, P - range. Dependences of the viscosity of model kimberlite and basaltic melts on temperature are consistent to the exponential Arrenian equation in the T, P - range of experimental study. The correct values of activation energies of viscous flow of kimberlite melts have been obtained for the first time. The activation energies of viscous flow of model kimberlite melts exponentially increase with increasing pressure and are equal: E = 130 ± 1.3 kJ/mole at moderate pressure (P = 100 MPa) and E = 160 ± 1.6 kJ/mole at high pressure (P = 5.5 GPa). It has been established too that the viscosity of model kimberlite melts exponentially increases on about half order of magnitude with increasing pressures from 100 MPa to 7.5 GPa at the isothermal condition (1800 oC). It has been established that viscosity of model kimberlite melts at the moderate pressure (100 MPa) is lover on about one order of magnitude to compare with the viscosity of basaltic melts, but at high pressure range (5.5 - 7.5 GPa), on the contrary, is higher on about half order of magnitude at the same values of the temperatures. Here we use both a new experimental data on viscosity of kimberlite melts and our structural chemical model for calculation and prediction the viscosity of magmatic melts [1] to determine the fundamental features of viscosity of

Russian military in the year 2000. Master's thesis

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

McIlmail, J.F.; Jaworski, J.L.

Through the use of content analysis, this paper attempts to paint a picture of the Russian military in the year 2000 and its impact on the US national security strategy. The research begins by defining the origin of Russian national security policy and how that translates into military policy and doctrine. A framework for evaluating Russian military doctrines is provided with a chronology of the military reform process and the related doctrinal reforms that has its birth in the 1987 announcement of a defensive-defense. Following from the doctrinal variant framework the new strategic missions of the 1992 draft military doctrinemore » are presented with an analysis that shows they are a clear departure from the past and truly represent a defensive-defense type doctrine. Additionally, a comparison is made with the current military reform ongoing in Russia with the historical precedent of the Russian military reform of 1924-25. A rough outline of the separate branches of the Russian military both present and future is provided based on the ongoing trends in the reform process. This thumbnail sketch of the Russian military then assists in the analysis and conclusion that even after a possible 50% cutback in US military spending, in the year 2000 the Russian military will not pose a threat to US national security. The major caveat to this conclusion is in the realm of nuclear weapons and this issue is therefore discussed in some length....Former Soviet Union, Russia, Commonwealth of Independent States (CIS), Conventional Forces in Europe treaty (CFE), European Security, New National Security Strategy, Nuclear Weapons, Nuclear Strategy.« less

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

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.

Phonotactic constraints: Implications for models of oral reading in Russian.

PubMed

Ulicheva, Anastasia; Coltheart, Max; Saunders, Steven; Perry, Conrad

2016-04-01

The present article investigates how phonotactic rules constrain oral reading in the Russian language. The pronunciation of letters in Russian is regular and consistent, but it is subject to substantial phonotactic influence: the position of a phoneme and its phonological context within a word can alter its pronunciation. In Part 1 of the article, we analyze the orthography-to-phonology and phonology-to-phonology (i.e., phonotactic) relationships in Russian monosyllabic words. In Part 2 of the article, we report empirical data from an oral word reading task that show an effect of phonotactic dependencies on skilled reading in Russian: humans are slower when reading words where letter-phoneme correspondences are highly constrained by phonotactic rules compared with those where there are few or no such constraints present. A further question of interest in this article is how computational models of oral reading deal with the phonotactics of the Russian language. To answer this question, in Part 3, we report simulations from the Russian dual-route cascaded model (DRC) and the Russian connectionist dual-process model (CDP++) and assess the performance of the 2 models by testing them against human data. (c) 2016 APA, all rights reserved).

Climate change and zoonotic infections in the Russian Arctic

PubMed Central

Revich, Boris; Tokarevich, Nikolai; Parkinson, Alan J.

2012-01-01

Climate change in the Russian Arctic is more pronounced than in any other part of the country. Between 1955 and 2000, the annual average air temperature in the Russian North increased by 1.2°C. During the same period, the mean temperature of upper layer of permafrost increased by 3°C. Climate change in Russian Arctic increases the risks of the emergence of zoonotic infectious diseases. This review presents data on morbidity rates among people, domestic animals and wildlife in the Russian Arctic, focusing on the potential climate related emergence of such diseases as tick-borne encephalitis, tularemia, brucellosis, leptospirosis, rabies, and anthrax. PMID:22868189

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

Rusistika: The Russian Journal of the Association for Language Learning, 1994-1997.

ERIC Educational Resources Information Center

Tejerizo, Margaret, Ed.

1997-01-01

This journal focuses on the teaching and learning of Russian. Selected articles include the following: "Soviet Cinema: Women's films"; "Learning and Teaching Russian"; "Russian in a Weekend?""Words Having a Religious Connotation in Russian"; "Grammar and Communication: The Ab Initio Russian…

Presence of Russian honey bee genotypes in swarms in Louisiana.

USDA-ARS?s Scientific Manuscript database

Swarm traps were placed in an area around USDA, ARS apiaries near Baton Rouge, Louisiana, which had contained ARS Russian and other honey bees for several years. Eighty swarms were sampled and analyzed for their genotype (Russian, hybrid or non-Russian) and mite infestation percentages. Ten swarms...

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.

Does Russian Democracy Have a Future?

DTIC Science & Technology

1994-01-01

Zhirinovsky, the Turkic specialist and child of Kazakhstan, also has called upon Russia to expand to the south. This "final thrust to the south" Zhirinovsky...Russian nationalists, let alone the Narod, had neither anticipated nor were psychologically prepared to face. Even the various referenda in Ukraine...nationalists perceived as the historic "lands of Rus" was an event for which the Russian body politic was not psychologically prepared, and initially

Russian Anti-Americanism: Origins and Implications

DTIC Science & Technology

2008-09-01

a growing pool of crime , inadequate social protection, and emptiness.29 So what does this mean? While Russians do not necessarily want to become... crime and corruption. Subconsciously driven by a strong Soviet upbringing, Putin has arguably transformed himself into a Russian post-Soviet tsar. He is...of Russia, Official Portal,17 February 2008, http://kremlin.ru/eng/text/ speeches /2008/02/14/1011_type82915_160266.shtml (accessed 12 March 2008

The "Politicizing" of Russian History Education in the Russian Media

ERIC Educational Resources Information Center

Zajda, Joseph

2014-01-01

This article examines the current debate on the politics of history education reforms and the new history textbooks for secondary schools in the Russian Federation. Recent reforms in history education, standards and prescribed history textbooks by the Ministry of Education and Science demonstrate a pronounced ideological shift in the national…

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.)

Communicative, Educational, Pedagogical Objectives and Planning in Russian Language Teaching

ERIC Educational Resources Information Center

Evtyugina, Alla A.; Hasanova, Irina I.; Kotova, Svetlana S.; Sokolova, Anastasia N.; Svetkina, Irina A.

2016-01-01

The relevance of the problem stems from the necessity to distinctly plan educational process and set the goals for successful mastering of Russian language by foreign students in Russian higher educational institutions. The article is aimed at defining the foreign students' objectives for Russian language training, allowing them to get involved…

Recent Anthologies of Eighteenth-Century Russian Literature: A Review Article.

ERIC Educational Resources Information Center

Edgerton, William B.

1968-01-01

Harold Segel's recently published anthology of eighteenth-century Russian literature in English is compared with the Soviet anthologies of Gukovskij and Kokorev (in Russian), the Polish anthology of Jakubowski (in Russian with Polish notes), and the early nineteenth-century Wiener anthology (in English). All of these works are described in some…

Russian Civic Education and Social Studies Education at Purdue University

ERIC Educational Resources Information Center

Nelson, Lynn R.; Rapoport, Anatoli

2005-01-01

There are three distinct Russian civic education programs at Purdue University (the Civics Mosaic program, the Training of Professors for Civic Education program, and the Russian Maymester program). The programs are loosely joined together by a civic education purpose, but remain distinct in the nature of their participants--Russian and American…

The Tripartite Seminars at the Russian Methodological and Training Center: An Overview of Progress in Nuclear Materials Control and Accounting in the Russian Federation

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

Pshakin, Gennady; Dickman, Deborah A.; Ryazanov, Boris

2003-09-16

The development of the Russian Methodological and Training Center (RMTC) has been an important element of collaboration between the Russian Federation (RF), the European Union and the US. The RMTC, located at the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russia has been designated by the Russian Ministry of Atomic Energy (Minatom) to provide nuclear materials protection, control and Accounting training to Minatom and the Federal Nuclear and Radiation Safety Authority (Gosatomnadzor) personnel. In addition, the RMTC was chartered with the responsibility of providing a venue for exchange of experiences and information between Russian and international specialists.

Thermal Analysis on the Shipment of Russian Plutonium Fuel

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

Or, Chuen T; Skrabek, Emanuel A; Carpenter, Robert T

Paper presented at the 12th Symposium on Space Nuclear Power and Propulsion in Albuquerque, NM in January 1995. The Mound 9516 shipping package was designed for the shipment of Plutonium-238 fuel. One of the shipping configurations is the Russian Pu-238 powder can. Computer models using SINDA were created to predict the temperatures of the package under normal conditions of transport (NCT: 38oC ambient temperature), under hypothetical accident conditions (HAC: engulfed in fire for 30 minutes), and inside a standard cargo container. Pressure increases inside the package due to the expansion of the trapped gases and helium gas generation from isotopemore » decay were also analyzed. There is a duplicate copy and also a copy in the ESD Files.« less

27 CFR 9.66 - Russian River Valley.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2014-04-01 2014-04-01 false Russian River Valley. 9.66 Section 9.66 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY ALCOHOL AMERICAN VITICULTURAL AREAS Approved American Viticultural Areas § 9.66 Russian River Valley. (a) Name. The name of the...

27 CFR 9.66 - Russian River Valley.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2012-04-01 2012-04-01 false Russian River Valley. 9.66 Section 9.66 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY LIQUORS AMERICAN VITICULTURAL AREAS Approved American Viticultural Areas § 9.66 Russian River Valley. (a) Name. The name of the...

27 CFR 9.66 - Russian River Valley.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2011-04-01 2011-04-01 false Russian River Valley. 9.66 Section 9.66 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY LIQUORS AMERICAN VITICULTURAL AREAS Approved American Viticultural Areas § 9.66 Russian River Valley. (a) Name. The name of the...

27 CFR 9.66 - Russian River Valley.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2013-04-01 2013-04-01 false Russian River Valley. 9.66 Section 9.66 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY ALCOHOL AMERICAN VITICULTURAL AREAS Approved American Viticultural Areas § 9.66 Russian River Valley. (a) Name. The name of the...

27 CFR 9.66 - Russian River Valley.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Russian River Valley. 9.66 Section 9.66 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY LIQUORS AMERICAN VITICULTURAL AREAS Approved American Viticultural Areas § 9.66 Russian River Valley. (a) Name. The name of the...

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

Russian Soyuz in Launch Position

NASA Technical Reports Server (NTRS)

2000-01-01

The Soyuz TM-31 launch vehicle is shown in the vertical position for its launch from Baikonur, carrying the first resident crew to the International Space Station. The Russian Soyuz launch vehicle is an expendable spacecraft that evolved out of the original Class A (Sputnik). From the early 1960s until today, the Soyuz launch vehicle has been the backbone of Russia's marned and unmanned space launch fleet. Today, the Soyuz launch vehicle is marketed internationally by a joint Russian/French consortium called STARSEM. As of August 2001, there have been ten Soyuz missions under the STARSEM banner.

Current knowledge and attitudes: Russian olive biology, ecology and management

Treesearch

Sharlene E. Sing; Kevin J. Delaney

2016-01-01

The primary goals of a two-day Russian olive symposium held in February 2014 were to disseminate current knowledge and identify data gaps regarding Russian olive biology and ecology, distributions, integrated management, and to ascertain the feasibility and acceptance of a proposed program for classical biological control of Russian olive. The symposium was...

Ten Years in the Life of Russian Libraries.

ERIC Educational Resources Information Center

Greening, Joyce Martin

1995-01-01

Examines the effects of the political events of the last decade on Russian libraries. Topics include libraries in the old Soviet Union, decentralization of control and financing, relaxation of censorship, changes in the Russian publishing industry, channels of distribution, and future needs. (LRW)

Hereditary cataracts in Russian Blue cats.

PubMed

Nygren, Karin; Jalomäki, Sari; Karlstam, Lena; Narfström, Kristina

2018-01-01

Objectives The purpose of this study was to investigate the prevalence of cataracts in the Russian Blue breed of cats in Sweden, and to describe the clinical appearance of this presumed inherited form of cataract. Methods A total of 66 Russian Blue cats were examined in Sweden, between March and October 2014, using standard examination techniques. The examined cats were between 3 months and 14 years of age. Pedigrees were collected from all examined cats for genetic studies. Results Mild-to-severe forms of mainly bilateral cataracts were observed in 22/66 examined Russian Blue cats of both sexes. Two affected cats were <1 year of age. The most frequently observed appearance of a cataract was a small triangular, Y-shaped or circular opacity at the border of the posterior nucleus and the anterior part of the posterior cortex, which caused no observable visual impairment. More extended forms were observed in 6/22 cats, with involvement of both the nucleus and either the entire cortex or parts of the posterior and/or anterior cortex. Visual impairment or blindness was observed in the latter six cases. Pedigree analyses indicated a simple autosomal recessive mode of inheritance for the defect, although a dominant mode with incomplete penetrance could not be excluded. Conclusions and relevance This study indicates that the Russian Blue breed of cat is affected by hereditary cataracts. The high prevalence in young cats and the characteristic location of the most frequently observed defect in the study suggest an early onset type of cataract. The breeders should be aware of this defect and have their cats examined by a veterinary ophthalmologist before breeding of an individual Russian Blue cat is considered.

Bibliometric Indicators of Russian Journals by JCR-Science Edition, 1995-2010

PubMed Central

Libkind, A.N.; Markusova, V.A.; Mindeli, L.E.

2013-01-01

A representative empirical bibliometric analysis of Russian journals included in the Journal Citation Reports-Science Edition (JCR-SE) for the time period 1995–2010 was conducted at the macro level (excluding the subject categories). It was found that the growth in the number of articles covered by JCR (a 1.8-fold increase compared to 1995) is ahead of the growth rates of Russian publications (1.2-fold increase). Hence, the share of Russian articles covered by JCR-SE was down from 2.5% in 1995 to 1.7% in 2010. It was determined that the number of articles published in an average Russian journal reduced by 20% as compared to the number of articles in an average journal of the full data set. These facts could partly shed light on the question why Russian research performance is staggering (approximately 30,000 articles per year), although the coverage of Russian journals has expanded to 150 titles. Over the past 15 years, a twofold increase in the impact factor of the Russian journals has been observed, which is higher than that for the full data set of journals (a 1.4-fold increase). Measures to improve the quality of Russian journals are proposed. PMID:24303198

Bibliometric Indicators of Russian Journals by JCR-Science Edition, 1995-2010.

PubMed

Libkind, A N; Markusova, V A; Mindeli, L E

2013-07-01

A representative empirical bibliometric analysis of Russian journals included in the Journal Citation Reports-Science Edition (JCR-SE) for the time period 1995-2010 was conducted at the macro level (excluding the subject categories). It was found that the growth in the number of articles covered by JCR (a 1.8-fold increase compared to 1995) is ahead of the growth rates of Russian publications (1.2-fold increase). Hence, the share of Russian articles covered by JCR-SE was down from 2.5% in 1995 to 1.7% in 2010. It was determined that the number of articles published in an average Russian journal reduced by 20% as compared to the number of articles in an average journal of the full data set. These facts could partly shed light on the question why Russian research performance is staggering (approximately 30,000 articles per year), although the coverage of Russian journals has expanded to 150 titles. Over the past 15 years, a twofold increase in the impact factor of the Russian journals has been observed, which is higher than that for the full data set of journals (a 1.4-fold increase). Measures to improve the quality of Russian journals are proposed.

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

Indissoluble Connection of Russian MC&A System Sustainability with that of the Russian Methodological & Training Center

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

Ryazanov, Boris G.; Goryunov, Victor; Pshakin, Gennady M.

During the past 10 years of collaboration under the U.S.-Russian Materials Protection, Control and Accounting (MPC&A) Program great efforts were made to transform and upgrade the Russian Federal Nuclear Materials Control and Accounting (MC&A) System. The efforts were focused not only on MC&A systems for nuclear facilities but also on creating and developing the system infrastructure, including legislation, state inspection and agency monitoring, training of operators and inspectors, instrument and methodological support, and the Federal Information System (FIS). At present the most important and urgent goal is to provide sustainability of MC&A systems at the existing level or at themore » level that will be achieved in 2007-2008. Since the very beginning of the program, the Russian Methodological and Training Center (RMTC) activities have been focused on intensive training of the personnel as well as the methodological support necessary for transformation and development of the entire system and its elements located at nuclear facilities. Sustainability of the federal MC&A system is impossible without advanced training of personnel and methodological support for upgrading of system elements at nuclear facilities. That is why the RMTC sustainability is one of the key conditions required for the system sustainability as a whole. The paper presents the results of analysis of the conditions for the Russian MC&A system sustainable development in conjunction with the RMTC sustainability.« less

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

"Signs of honor" among Russian inmates in Israel's prisons.

PubMed

Shoham, Efrat

2010-12-01

The unique nature of Israeli society as an immigrant society has also affected the prison population in Israel. This article focuses on a social and cultural phenomenon that particularly characterizes the prisoners of Russian origin, the phenomenon of tattoos. Using postmodernist theories, the article examines the function of the tattoo among Russian prisoners and the role it plays in constructing the criminal self-identity of these inmates in Israeli prisons. The tattoos observed during 2005-2006 among the Russian prisoners in four major Israeli prisons reflect the values of the Russian criminal subculture from which they evolved and were imported. This subculture is characterized by a hierarchical class structure and manifestations of machismo, domination, defiance, rebellion, and open antagonism against the Establishment and its representatives.

Proceedings of the 5. joint Russian-American computational mathematics conference

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

NONE

1997-12-31

These proceedings contain a record of the talks presented and papers submitted by participants. The conference participants represented three institutions from the United States, Sandia National Laboratories (SNL), Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and two from Russia, Russian Federal Nuclear Center--All Russian Research Institute of Experimental Physics (RFNC-VNIIEF/Arzamas-16), and Russian Federal Nuclear Center--All Russian Research Institute of Technical Physics (RFNC-VNIITF/Chelyabinsk-70). The presentations and papers cover a wide range of applications from radiation transport to materials. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

Evaluation of Pressurization Fatigue Life of 1441 Al-li Fuselage Panel

NASA Technical Reports Server (NTRS)

Bird, R. Keith; Dicus, Dennis I.; Fridlyander, Joseph; Davydov, Valentin

1999-01-01

A study was conducted to evaluate the pressurization fatigue life of fuselage panels with skins fabricated from 1441 Al-Li, an attractive new Russian alloy. The study indicated that 1441 Al-Li has several advantages over conventional aluminum fuselage skin alloy with respect to fatigue behavior. Smooth 1441 Al-Li sheet specimens exhibited a fatigue endurance limit similar to that for 1163 Al (Russian version of 2024 Al) sheet. Notched 1441 Al-Li sheet specimens exhibited greater fatigue strength and longer fatigue life than 1163 Al. In addition, Tu-204 fuselage panels fabricated by Tupolev Design Bureau using Al-Li skin and ring frames with riveted 7000-series aluminum stiffeners had longer pressurization fatigue lives than did panels constructed from conventional aluminum alloys. Taking into account the lower density of this alloy, the results suggest that 1441 Al-Li has the potential to improve fuselage performance while decreasing structural weight.

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

Illness perception differences between Russian- and Hebrew-speaking Israeli oncology patients.

PubMed

Popov, Nadia; Heruti, Irit; Levy, Sigal; Lulav-Grinwald, Doron; Bar-Sela, Gil

2014-03-01

Illness perception influences health and illness behaviors. This study was designed to estimate illness perception differences between Russian-speaking and Hebrew-speaking Israeli oncology patients. Changes in illness perception associated with time spent in Israel among Russian-speaking patients were also evaluated. Additionally, we evaluated differences in illness perception of patients exposed to Chernobyl's consequences. A total of 144 oncology patients (77 Hebrew-speaking, 67 Russian-speaking) completed personal data questionnaires and The illness perception questionnaire revised, translated into Russian for this study. Significantly more Russian-speaking oncology patients perceived their illness as chronic and having negative consequences on life (p < .01). Russian-speaking oncology patients tend to have a more negative perception of cancer compared to Hebrew-speaking patients. Time spent in Israel may create more positive perceptions of cancer among these patients. No illness perception differences were found concerning Chernobyl consequences.

Teaching Russian Civilization.

ERIC Educational Resources Information Center

Vakar, Nicholas

1949-01-01

Prior to World War Two, Slavic studies in America treated history, literature, and language as isolated disciplines and often neglected the study of Russian literature written after 1917. The pragmatic necessities of the war questioned the relevance of this traditional approach and specialists appeared, concentrating their efforts on the recent…

Russian-US collaboration on implementation of the active well coincidence counter (AWCC)

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

Mozhajev, V.; Pshakin, G.; Stewart, J.

The feasibility of using a standard AWCC at the Obninsk IPPE has been demonstrated through active measurements of single UO{sub 2} (36% enriched) disks and through passive measurements of plutonium metal disks used for simulating reactor cores. The role of the measurements is to verify passport values assigned to the disks by the facility, and thereby facilitate the mass accountability procedures developed for the very large inventory of fuel disks at the facility. The AWCC is a very flexible instrument for verification measurements of the large variety of nuclear material items at the Obninsk IPPE and other Russian facilities. Futuremore » work at the IPPE will include calibration and verification measurements for other materials, both in individual disks and in multi-disk storage tubes; it will also include training in the use of the AWCC.« less

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

Urinary nickel excretion in populations living in the proximity of two russian nickel refineries: a Norwegian-Russian population-based study.

PubMed Central

Smith-Sivertsen, T; Tchachtchine, V; Lund, E; Bykov, V; Thomassen, Y; Norseth, T

1998-01-01

The Russian nickel refineries located in the cities of Nikel and Zapolyarny close to the Norwegian border are responsible for extensive sulfur dioxide and nickel pollution, as well as severe ecological damage in both countries. The aim of our study was to investigate human nickel exposure in the populations living on both sides of the Norwegian-Russian border. The design was a cross-sectional population-based study of adults aged 18-69 years residing in Sor-Varanger municipality, Norway, and Nikel and Zapolyarny, Russia, during 1994 and 1995. Individual exposure to nickel was assessed by measurements of nickel in urine using electrothermal atomic absorption spectrometry. For controls, urine was collected from adults in the Russian cities of Apatity and Umba (Kola Peninsula) and the Norwegian city of Tromso, all of which are locations without nearby point sources of nickel. Altogether 2,233 urine specimens were analysed for nickel. People living in Nikel had the highest concentrations (median 3.4 microg/l), followed by Umba (median 2.7 microg/l), Zapolyarny (median 2.0 microg/l), Apatity (median 1.9 microg/l), Tromso (median 1.2 microg/l), and Sor-Varanger (median 0.6 microg/l). Regardless of geographical location, the Russian study groups all had a higher urinary-nickel average than those in Norway (p<0.001). With the exception of Nikel, neither the Russian nor the Norwegian urinary-nickel levels were associated with residence location near a Russian nickel refinery. We concluded that industrial nickel pollution alone could not explain the observed discrepancy between Norway and Russia; we also discuss other possible nickel exposure sources that may account for the high urinary levels found in Russia. Images Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 PMID:9681979

Russian Planetary Program: Phobos and the Moon

NASA Astrophysics Data System (ADS)

Galimov, E. M.; Marov, M. Ya.; Politshuk, G. M.; Zeleniy, L. M.

2006-08-01

Planetary exploration is a cornerstone of space science and technology development. Russia has a great legacy of the world recognized former space missions to the Moon and planets. Strategy of the Russian Federal Space Agency and the Russian Academy of Sciences planetary program for the coming decade is focused on space vehicle of new generation. The basic concept of this spacecraft development is the modern technology utilization, significant cost reduction and meeting objectives of the important science return. The bottom line is the use of middle class Soyuz-type launcher, which places the principal constraint on mass of the vehicle and mission profile. Flexibility in the design of space vehicle, including a possibility of SEP technology utilization, facilitates its adaptability for extended program of the solar system exploration. As the first step, the project is optimized around sample return mission from satellite of Mars Phobos ("Phobos-Grunt" or PSR) which is in the list of the Russian Federal Space Program for 2006 to 2015. It is to be launched in 2009 and completed in 2012. The experience gained from the former Russian "Phobos 88" serves as a clue to provide an important basis for the mission concept enabling solution of many problems of the project design and its implementation. There is a challenge to return relic matter from such small body like Phobos for the ground labs comprehensive study. The payload is also targeted for in-flight and extended remote sensing and in situ measurements using the capable instrument packages. The project is addressed as a milestone in the Russian program of the solar system study, with a potential for future ambitious missions to asteroids and comets pooling international efforts. Also endorsed by the Russian Federal Space Program is "Luna-Glob" mission to the Moon tentatively scheduled for 2011. The goal is to advance lunar science with the well instrumented orbiter, lander, and the network of penetrators. Return back

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.

Social Factors Influencing Russian Male Alcohol Use over the Life Course: A Qualitative Study Investigating Age Based Social Norms, Masculinity, and Workplace Context

PubMed Central

Keenan, Katherine; Saburova, Lyudmila; Bobrova, Natalia; Elbourne, Diana; Ashwin, Sarah; Leon, David A.

2015-01-01

The massive fluctuations occurring in Russian alcohol-related mortality since the mid-1980s cannot be seen outside of the context of great social and economic change. There is a dearth of qualitative studies about Russian male drinking and especially needed are those that address social processes and individual changes in drinking. Conducted as part of a longitudinal study on men’s alcohol consumption in Izhevsk, this qualitative study uses 25 semi-structured biographical interviews with men aged 33–60 years to explore life course variation in drinking. The dominant pattern was decreasing binge and frequent drinking as men reached middle age which was precipitated by family building, reductions in drinking with work colleagues, and health concerns. A minority of men described chaotic drinking histories with periods of abstinence and heavy drinking. The results highlight the importance of the blue-collar work environment for conditioning male heavy drinking in young adulthood through a variety of social, normative and structural mechanisms. Post-Soviet changes had a structural influence on the propensity for workplace drinking but the important social function of male drinking sessions remained. Bonding with workmates through heavy drinking was seen as an unavoidable and essential part of young men’s social life. With age peer pressure to drink decreased and the need to perform the role of responsible breadwinner put different behavioural demands on men. For some resisting social pressure to drink became an important site of self-determination and a mark of masculine maturity. Over the lifetime the place where masculine identity was asserted shifted from the workplace to the home, which commonly resulted in a reduction in drinking. We contribute to existing theories of Russian male drinking by showing that the performance of age-related social roles influences Russian men’s drinking patterns, drinking contexts and their attitudes. Further research should be

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

LESSONS FROM THE FRONT: A CASE STUDY OF RUSSIAN CYBER WARFARE

DTIC Science & Technology

2015-12-01

Lessons From The Front: A Case Study Of Russian Cyber Warfare looks to capitalize on the lessons learned from the alleged Russian cyber-offensive on...through the careful analysis and comparison of two disparate conflicts related by their collision with Russian cyber - warfare . Following case study

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.

Palatalization and Intrinsic Prosodic Vowel Features in Russian

ERIC Educational Resources Information Center

Ordin, Mikhail

2011-01-01

The presented study is aimed at investigating the interaction of palatalization and intrinsic prosodic features of the vowel in CVC (consonant+vowel+consonant) syllables in Russian. The universal nature of intrinsic prosodic vowel features was confirmed with the data from the Russian language. It was found that palatalization of the consonants…

Russian Foreign Policy in Historical and Current Context: A Reassessment

DTIC Science & Technology

2015-01-01

concerns did not slow the growth of Russian trade with the European Union or Russian tourism to Western European countries. Moreover, before the...Russia’s economic interests. Crimea’s primary economic activities have been tourism and hosting the Russian (and Ukrainian) Black Sea fleets. As incomes...age benefits.41 Following the Beslan hostage crisis, elected gover - nors were replaced by appointed representatives. (And although this was

Russian Political, Economic, and Security Issues and U.S. Interests

DTIC Science & Technology

2010-01-29

potential.77 Russian Restrictions on Meat Imports Russia has been a very important market for U.S. pork and poultry producers. At the end of 2009...34 Russian Restrictions on Meat Imports ........................................................................... 36 U.S. Assistance to Russia...and made progress in respecting the rule of law and human rights. Russia’s restrictions on meat imports are becoming a major irritant in U.S.-Russian

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

Russian Business Schools in a Time of Transition

ERIC Educational Resources Information Center

Mechitov, Alexander I.; Moshkovich, Helen M.

2006-01-01

In this study, the authors reviewed the development of Russian business education in the past decade. This development, fueled by historic changes in Russian society, has affected all aspects of business education, including its organizational structures, demand in different business areas, and mode of teaching. In a short period of time, Russian…

The Sociocultural Factors of Russian Stagnation and Modernization

ERIC Educational Resources Information Center

Lapin, N. I.

2012-01-01

An analysis of the results of six Russian Longitudinal Monitoring Surveys, "The Values and Interests of the Population of Russia" (1990, 1994, 1998, 2002, 2006, and 2010), conducted by the Center for the Study of Sociocultural Changes at the Institute of Philosophy, Russian Academy of Sciences, provides evidence that two stages in the…

What the Russian School Ought to Be Like.

ERIC Educational Resources Information Center

Goncharov, I.

1995-01-01

Asserts that Russian society and Russian schools are going through a profound crisis. Maintains that the best approach to solving social and educational problems is to restore and develop national principles and group cohesion. Criticizes the United States and Japanese educational systems and calls for Russia to follow its own path. (CFR)

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