Langmuir probe study of a magnetically enhanced RF plasma source at pressures below 0.1 Pa

Science.gov (United States)

Kousal, Jaroslav; Tichý, Milan; Šebek, Ondřej; Čechvala, Juraj; Biederman, Hynek

2011-08-01

The majority of plasma polymerization sources operate at pressures higher than 1 Pa. At these pressures most common deposition methods do not show significant directionality. One way of enhancing the directional effects is to decrease the working pressure to increase the mean free path of the reactive molecules. The plasma source used in this work was designed to study the plasma polymerization process at pressures below 0.1 Pa. The source consists of the classical radio frequency (RF) (13.56 MHz, capacitive coupled) tubular reactor enhanced by an external magnetic circuit. The working gas is introduced into the discharge by a capillary. This forms a relatively localized zone of higher pressure where the monomer is activated. Due to the magnetic field, the plasma is constricted near the axis of the reactor with nearly collisionless gas flow. The plasma parameters were obtained using a double Langmuir probe. Plasma density in the range ni = 1013-1016 m-3 was obtained in various parts of the discharge under typical conditions. The presence of the magnetic field led to the presence of relatively strong electric fields (103 V m-1) and relatively high electron energies up to several tens of eV in the plasma.

Langmuir probe study of a magnetically enhanced RF plasma source at pressures below 0.1 Pa

Energy Technology Data Exchange (ETDEWEB)

Kousal, Jaroslav; Tichy, Milan; Sebek, Ondrej; Cechvala, Juraj; Biederman, Hynek, E-mail: jaroslav.kousal@mff.cuni.cz [Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, 180 00, Prague 8 (Czech Republic)

2011-08-15

The majority of plasma polymerization sources operate at pressures higher than 1 Pa. At these pressures most common deposition methods do not show significant directionality. One way of enhancing the directional effects is to decrease the working pressure to increase the mean free path of the reactive molecules. The plasma source used in this work was designed to study the plasma polymerization process at pressures below 0.1 Pa. The source consists of the classical radio frequency (RF) (13.56 MHz, capacitive coupled) tubular reactor enhanced by an external magnetic circuit. The working gas is introduced into the discharge by a capillary. This forms a relatively localized zone of higher pressure where the monomer is activated. Due to the magnetic field, the plasma is constricted near the axis of the reactor with nearly collisionless gas flow. The plasma parameters were obtained using a double Langmuir probe. Plasma density in the range n{sub i} = 10{sup 13}-10{sup 16} m{sup -3} was obtained in various parts of the discharge under typical conditions. The presence of the magnetic field led to the presence of relatively strong electric fields (10{sup 3} V m{sup -1}) and relatively high electron energies up to several tens of eV in the plasma.

Pressurized-water reactors

International Nuclear Information System (INIS)

Bush, S.H.

1983-03-01

An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

Reactor pressure boundary materials

International Nuclear Information System (INIS)

Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

2002-04-01

With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

Effects of gap and elevated pressure on ethanol reforming in a non-thermal plasma reactor

International Nuclear Information System (INIS)

Hoang, Trung Q; Zhu Xinli; Lobban, Lance L; Mallinson, Richard G

2011-01-01

Production of hydrogen for fuel cell vehicles, mobile power generators and for hydrogen-enhanced combustion from ethanol is demonstrated using energy-efficient non-thermal plasma reforming. A tubular reactor with a multipoint electrode system operated in pulsed mode was used. Complete conversion can be achieved with high selectivity (based on ethanol) of H 2 and CO of 111% and 78%, respectively, at atmospheric pressure. An elevated pressure of 15 psig shows improvement of selectivity of H 2 and CO to 120% and 87%, with a significant reduction of C 2 H x side products. H 2 selectivity increased to 127% when a high ratio (29.2) of water-to-ethanol feed was used. Increasing CO 2 selectivity is observed at higher water-to-ethanol ratios indicating that the water gas shift reaction occurs. A higher productivity and lower C 2 H x products were observed at larger gas gaps. The highest overall energy efficiency achieved, including electrical power consumption, was 82% for all products or 66% for H 2 only.

Destruction of Bacillus subtilis cells using an atmospheric-pressure dielectric capillary electrode discharge plasma

International Nuclear Information System (INIS)

Panikov, N.S.; Paduraru, S.; Crowe, R.; Ricatto, P.J.; Christodoulatos, C.; Becker, K.

2002-01-01

The results of experiments aimed at the investigation of the destruction of spore-forming bacteria, which are believed to be among the most resistant microorganisms, using a novel atmospheric-pressure dielectric capillary electrode discharge plasma are reported. Various well-characterized cultures of Bacillus subtilis were prepared, subjected to atmospheric-pressure plasma jets emanating from a plasma shower reactor operated either in He or in air (N 2 /O 2 mixture) at various power levels and exposure times, and analyzed after plasma treatment. Reductions in colony-forming units ranged from 10 4 (He plasma) to 10 8 (air plasma) for plasma exposure times of less than 10 minutes. (author)

Flow reactor studies of non-equilibrium plasma-assisted oxidation of n-alkanes.

Science.gov (United States)

Tsolas, Nicholas; Lee, Jong Guen; Yetter, Richard A

2015-08-13

The oxidation of n-alkanes (C1-C7) has been studied with and without the effects of a nanosecond, non-equilibrium plasma discharge at 1 atm pressure from 420 to 1250 K. Experiments have been performed under nearly isothermal conditions in a flow reactor, where reactive mixtures are diluted in Ar to minimize temperature changes from chemical reactions. Sample extraction performed at the exit of the reactor captures product and intermediate species and stores them in a multi-position valve for subsequent identification and quantification using gas chromatography. By fixing the flow rate in the reactor and varying the temperature, reactivity maps for the oxidation of fuels are achieved. Considering all the fuels studied, fuel consumption under the effects of the plasma is shown to have been enhanced significantly, particularly for the low-temperature regime (T<800 K). In fact, multiple transitions in the rates of fuel consumption are observed depending on fuel with the emergence of a negative-temperature-coefficient regime. For all fuels, the temperature for the transition into the high-temperature chemistry is lowered as a consequence of the plasma being able to increase the rate of fuel consumption. Using a phenomenological interpretation of the intermediate species formed, it can be shown that the active particles produced from the plasma enhance alkyl radical formation at all temperatures and enable low-temperature chain branching for fuels C3 and greater. The significance of this result demonstrates that the plasma provides an opportunity for low-temperature chain branching to occur at reduced pressures, which is typically observed at elevated pressures in thermal induced systems. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

Science.gov (United States)

Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

2015-12-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

International Nuclear Information System (INIS)

Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

2015-01-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

Energy Technology Data Exchange (ETDEWEB)

Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

2015-12-15

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

Generation of high-power-density atmospheric pressure plasma with liquid electrodes

International Nuclear Information System (INIS)

Dong Lifang; Mao Zhiguo; Yin Zengqian; Ran Junxia

2004-01-01

We present a method for generating atmospheric pressure plasma using a dielectric barrier discharge reactor with two liquid electrodes. Four distinct kinds of discharge, including stochastic filaments, regular square pattern, glow-like discharge, and Turing stripe pattern, are observed in argon with a flow rate of 9 slm. The electrical and optical characteristics of the device are investigated. Results show that high-power-density atmospheric pressure plasma with high duty ratio in space and time can be obtained. The influence of wall charges on discharge power and duty ratio has been discussed

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Real-time control of current and pressure profiles in tokamak plasmas

International Nuclear Information System (INIS)

Laborde, L.

2005-12-01

Recent progress in the field of 'advanced tokamak scenarios' prefigure the operation regime of a future thermonuclear fusion power plant. Compared to the reference regime, these scenarios offer a longer plasma confinement time thanks to increased magnetohydrodynamic stability and to a better particle and energy confinement through a reduction of plasma turbulence. This should give access to comparable fusion performances at reduced plasma current and could lead to a steady state fusion reactor since the plasma current could be entirely generated non-inductively. Access to this kind of regime is provided by the existence of an internal transport barrier, linked to the current profile evolution in the plasma, which leads to steep temperature and pressure profiles. The comparison between heat transport simulations and experiments allowed the nature of the barriers to be better understood as a region of strongly reduced turbulence. Thus, the control of this barrier in a stationary manner would be a remarkable progress, in particular in view of the experimental reactor ITER. The Tore Supra and JET tokamaks, based in France and in the United Kingdom, constitute ideal instruments for such experiments: the first one allows stationary plasmas to be maintained during several minutes whereas the second one provides unique fusion performances. In Tore Supra, real-time control experiments have been accomplished where the current profile width and the pressure profile gradient were controlled in a stationary manner using heating and current drive systems as actuators. In the JET tokamak, the determination of an empirical static model of the plasma allowed the current and pressure profiles to be simultaneously controlled and so an internal transport barrier to be sustained. Finally, the identification of a dynamic model of the plasma led to the definition of a new controller capable, in principle, of a more efficient control. (author)

Modeling of low pressure plasma sources for microelectronics fabrication

International Nuclear Information System (INIS)

Agarwal, Ankur; Bera, Kallol; Kenney, Jason; Rauf, Shahid; Likhanskii, Alexandre

2017-01-01

Chemically reactive plasmas operating in the 1 mTorr–10 Torr pressure range are widely used for thin film processing in the semiconductor industry. Plasma modeling has come to play an important role in the design of these plasma processing systems. A number of 3-dimensional (3D) fluid and hybrid plasma modeling examples are used to illustrate the role of computational investigations in design of plasma processing hardware for applications such as ion implantation, deposition, and etching. A model for a rectangular inductively coupled plasma (ICP) source is described, which is employed as an ion source for ion implantation. It is shown that gas pressure strongly influences ion flux uniformity, which is determined by the balance between the location of plasma production and diffusion. The effect of chamber dimensions on plasma uniformity in a rectangular capacitively coupled plasma (CCP) is examined using an electromagnetic plasma model. Due to high pressure and small gap in this system, plasma uniformity is found to be primarily determined by the electric field profile in the sheath/pre-sheath region. A 3D model is utilized to investigate the confinement properties of a mesh in a cylindrical CCP. Results highlight the role of hole topology and size on the formation of localized hot-spots. A 3D electromagnetic plasma model for a cylindrical ICP is used to study inductive versus capacitive power coupling and how placement of ground return wires influences it. Finally, a 3D hybrid plasma model for an electron beam generated magnetized plasma is used to understand the role of reactor geometry on plasma uniformity in the presence of E  ×  B drift. (paper)

Modeling of low pressure plasma sources for microelectronics fabrication

Science.gov (United States)

Agarwal, Ankur; Bera, Kallol; Kenney, Jason; Likhanskii, Alexandre; Rauf, Shahid

2017-10-01

Chemically reactive plasmas operating in the 1 mTorr-10 Torr pressure range are widely used for thin film processing in the semiconductor industry. Plasma modeling has come to play an important role in the design of these plasma processing systems. A number of 3-dimensional (3D) fluid and hybrid plasma modeling examples are used to illustrate the role of computational investigations in design of plasma processing hardware for applications such as ion implantation, deposition, and etching. A model for a rectangular inductively coupled plasma (ICP) source is described, which is employed as an ion source for ion implantation. It is shown that gas pressure strongly influences ion flux uniformity, which is determined by the balance between the location of plasma production and diffusion. The effect of chamber dimensions on plasma uniformity in a rectangular capacitively coupled plasma (CCP) is examined using an electromagnetic plasma model. Due to high pressure and small gap in this system, plasma uniformity is found to be primarily determined by the electric field profile in the sheath/pre-sheath region. A 3D model is utilized to investigate the confinement properties of a mesh in a cylindrical CCP. Results highlight the role of hole topology and size on the formation of localized hot-spots. A 3D electromagnetic plasma model for a cylindrical ICP is used to study inductive versus capacitive power coupling and how placement of ground return wires influences it. Finally, a 3D hybrid plasma model for an electron beam generated magnetized plasma is used to understand the role of reactor geometry on plasma uniformity in the presence of E  ×  B drift.

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Controlling hydrophilicity of polymer film by altering gas flow rate in atmospheric-pressure homogeneous plasma

International Nuclear Information System (INIS)

Kang, Woo Seok; Hur, Min; Lee, Jae-Ok; Song, Young-Hoon

2014-01-01

Graphical abstract: - Highlights: • Controlling hydrophilicity of polymer film by varying gas flow rate is proposed in atmospheric-pressure homogeneous plasma treatment. • Without employing additional reactive gas, requiring more plasma power and longer treatment time, hydrophilicity of polyimide films was improved after the low-gas-flow plasma treatment. • The gas flow rate affects the hydrophilic properties of polymer surface by changing the discharge atmosphere in the particular geometry of the reactor developed. • Low-gas-flow induced wettability control suggests effective and economical plasma treatment. - Abstract: This paper reports on controlling the hydrophilicity of polyimide films using atmospheric-pressure homogeneous plasmas by changing only the gas flow rate. The gas flow changed the discharge atmosphere by mixing the feed gas with ambient air because of the particular geometry of the reactor developed for the study, and a low gas flow rate was found to be favorable because it generated abundant nitrogen or oxygen species that served as sources of hydrophilic functional groups over the polymer surface. After low-gas-flow plasma treatment, the polymer surface exhibited hydrophilic characteristics with increased surface roughness and enhanced chemical properties owing to the surface addition of functional groups. Without adding any reactive gases or requiring high plasma power and longer treatment time, the developed reactor with low-gas-flow operation offered effective and economical wettability control of polyimide films

Characteristics of Atmospheric Pressure Rotating Gliding Arc Plasmas

Science.gov (United States)

Zhang, Hao; Zhu, Fengsen; Tu, Xin; Bo, Zheng; Cen, Kefa; Li, Xiaodong

2016-05-01

In this work, a novel direct current (DC) atmospheric pressure rotating gliding arc (RGA) plasma reactor has been developed for plasma-assisted chemical reactions. The influence of the gas composition and the gas flow rate on the arc dynamic behaviour and the formation of reactive species in the N2 and air gliding arc plasmas has been investigated by means of electrical signals, high speed photography, and optical emission spectroscopic diagnostics. Compared to conventional gliding arc reactors with knife-shaped electrodes which generally require a high flow rate (e.g., 10-20 L/min) to maintain a long arc length and reasonable plasma discharge zone, in this RGA system, a lower gas flow rate (e.g., 2 L/min) can also generate a larger effective plasma reaction zone with a longer arc length for chemical reactions. Two different motion patterns can be clearly observed in the N2 and air RGA plasmas. The time-resolved arc voltage signals show that three different arc dynamic modes, the arc restrike mode, takeover mode, and combined modes, can be clearly identified in the RGA plasmas. The occurrence of different motion and arc dynamic modes is strongly dependent on the composition of the working gas and gas flow rate. supported by National Natural Science Foundation of China (No. 51576174), the Specialized Research Fund for the Doctoral Program of Higher Education of China (No. 20120101110099) and the Fundamental Research Funds for the Central Universities (No. 2015FZA4011)

Plasma-gun fueling for tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.

1980-11-01

In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

Some aspects of reactor pressure vessel integrity

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic, G.J.

1996-01-01

Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

Seed disinfection effect of atmospheric pressure plasma and low pressure plasma on Rhizoctonia solani.

Science.gov (United States)

Nishioka, Terumi; Takai, Yuichiro; Kawaradani, Mitsuo; Okada, Kiyotsugu; Tanimoto, Hideo; Misawa, Tatsuya; Kusakari, Shinichi

2014-01-01

Gas plasma generated and applied under two different systems, atmospheric pressure plasma and low pressure plasma, was used to investigate the inactivation efficacy on the seedborne pathogenic fungus, Rhizoctonia solani, which had been artificially introduced to brassicaceous seeds. Treatment with atmospheric plasma for 10 min markedly reduced the R. solani survival rate from 100% to 3% but delayed seed germination. The low pressure plasma treatment reduced the fungal survival rate from 83% to 1.7% after 10 min and the inactivation effect was dependent on the treatment time. The seed germination rate after treatment with the low pressure plasma was not significantly different from that of untreated seeds. The air temperature around the seeds in the low pressure system was lower than that of the atmospheric system. These results suggested that gas plasma treatment under low pressure could be effective in disinfecting the seeds without damaging them.

Structural stability analysis considerations in fusion reactor plasma chamber design

International Nuclear Information System (INIS)

Delaney, M.J.; Cramer, B.A.

1978-01-01

This paper presents an approach to analyzing a toroidal plasma chamber for the prevention of both static and dynamic buckling. Results of stability analyses performed for the doublet shaped plasma chamber of the General Atomic 3.8 meter radius TNS ignition test reactor are presented. Load conditions are the static external atmospheric pressure load and the dynamic plasma disruption pulse load. Methods for analysis of plasma chamber structures are presented for both types of load. Analysis for static buckling is based on idealizing the plasma chamber into standard structural shapes and applying classical cylinder and circular torus buckling equations. Results are verified using the Buckling of Shells of Revolution (BOSOR4) finite difference computer code. Analysis for the dynamic loading is based on a pulse buckling analysis method for circular cylinders

Plasma behaviour in large reversed-field pinches and reactors

International Nuclear Information System (INIS)

Christiansen, J.P.; Bodin, H.A.B.; Carolan, P.G.; Johnston, J.W.; Newton, A.A.; Roberts, K.V.; Robinson, D.C.; Watts, M.R.C.; Piotrowicz, V.A.

1981-01-01

Recent analytic and numerical results on large reversed-field-pinch (RFP) systems and RFP reactors are presented. Predictions are made of the plasma behaviour in Eta Beta II, HBTXIA (under construction) and RFX (planned). The setting-up phase of an RFP is studied by using turbulence theory in transport equilibrium calculations, and estimates are made of the volt-seconds consumption for four different modes of field control. A prescription is given for a dynamo producing self-reversal which yields finite-β configurations. Residual instabilities of these equilibria may be resistive pressure-driven g-modes, and a new study of these modes that includes parallel viscosity indicates stability for anti β approximately 10%. The sustainment phase of the RFP is examined with tokamak scaling laws assumed for the energy confinement time. Temperatures in excess of 1keV are predicted for currents of 2MA in RFX. An operating cycle for a pulsed RFP reactor including gas puffing to reach ignition is proposed following a study of the energy replacement time for an Ohmically heated plasma. The scaling of the reactor parameters with minor radius is also investigated. (author)

Pressure releasing device for reactor container

International Nuclear Information System (INIS)

Takeda, Mika.

1994-01-01

In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

Reactor pressure vessel design

International Nuclear Information System (INIS)

Foehl, J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Calculation of gas-flow in plasma reactor for carbon partial oxidation

Science.gov (United States)

Bespala, Evgeny; Myshkin, Vyacheslav; Novoselov, Ivan; Pavliuk, Alexander; Makarevich, Semen; Bespala, Yuliya

2018-03-01

The paper discusses isotopic effects at carbon oxidation in low temperature non-equilibrium plasma at constant magnetic field. There is described routine of experiment and defined optimal parameters ensuring maximum enrichment factor at given electrophysical, gas-dynamic, and thermodymanical parameters. It has been demonstrated that at high-frequency generator capacity of 4 kW, supply frequency of 27 MHz and field density of 44 mT the concentration of paramagnetic heavy nuclei 13C in gaseous phase increases up to 1.78 % compared to 1.11 % for natural concentration. Authors explain isotopic effect decrease during plasmachemical separation induced by mixing gas flows enriched in different isotopes at the lack of product quench. With the help of modeling the motion of gas flows inside the plasma-chemical reactor based on numerical calculation of Navier-Stokes equation authors determine zones of gas mixing and cooling speed. To increase isotopic effects and proportion of 13C in gaseous phase it has been proposed to use quench in the form of Laval nozzle of refractory steel. The article represents results on calculation of optimal Laval Nozzle parameters for plasma-chemical reactor of chosen geometry of. There are also given dependences of quench time of products on pressure at the diffuser output and on critical section diameter. Authors determine the location of quench inside the plasma-chemical reactor in the paper.

Applications of plasma core reactors to terrestrial energy systems

International Nuclear Information System (INIS)

Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

International Nuclear Information System (INIS)

Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

2008-01-01

The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

Fundamentals of pressurized water reactors

International Nuclear Information System (INIS)

Murray, L.

1982-01-01

In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

Atmospheric-pressure plasma jet

Science.gov (United States)

Selwyn, Gary S.

1999-01-01

Atmospheric-pressure plasma jet. A .gamma.-mode, resonant-cavity plasma discharge that can be operated at atmospheric pressure and near room temperature using 13.56 MHz rf power is described. Unlike plasma torches, the discharge produces a gas-phase effluent no hotter than 250.degree. C. at an applied power of about 300 W, and shows distinct non-thermal characteristics. In the simplest design, two concentric cylindrical electrodes are employed to generate a plasma in the annular region therebetween. A "jet" of long-lived metastable and reactive species that are capable of rapidly cleaning or etching metals and other materials is generated which extends up to 8 in. beyond the open end of the electrodes. Films and coatings may also be removed by these species. Arcing is prevented in the apparatus by using gas mixtures containing He, which limits ionization, by using high flow velocities, and by properly shaping the rf-powered electrode. Because of the atmospheric pressure operation, no ions survive for a sufficiently long distance beyond the active plasma discharge to bombard a workpiece, unlike low-pressure plasma sources and conventional plasma processing methods.

Effect of plasma colloid osmotic pressure on intraocular pressure during haemodialysis

OpenAIRE

Tokuyama, T.; Ikeda, T.; Sato, K.

1998-01-01

BACKGROUND—In a previous case report, it was shown that an increase in plasma colloid osmotic pressure induced by the removal of fluid during haemodialysis was instrumental in decreasing intraocular pressure. The relation between changes in intraocular pressure, plasma osmolarity, plasma colloid osmotic pressure, and body weight before and after haemodialysis is evaluated.
METHODS—Intraocular pressure, plasma osmolarity, plasma colloid osmotic pressure, and body weight were evaluated before a...

Plasma engineering innovations for the ORNL TNS reactor

International Nuclear Information System (INIS)

Peng, Y.K.M.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

1977-01-01

Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma β values up to 10 percent and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1975-01-01

Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

Atmospheric pressure plasmas for surface modification of flexible and printed electronic devices: A review

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyong Nam; Lee, Seung Min; Mishra, Anurag [Department of Materials Science and Engineering, Sungkyunkwan University, Suwon, Gyeonggi-do 440-746 (Korea, Republic of); Yeom, Geun Young, E-mail: gyyeom@skku.edu [Department of Materials Science and Engineering, Sungkyunkwan University, Suwon, Gyeonggi-do 440-746 (Korea, Republic of); SKKU Advanced Institute of Nano Technology (SAINT), Sungkyunkwan University, Suwon, Gyeonggi-do 440-746 (Korea, Republic of)

2016-01-01

Recently, non-equilibrium atmospheric pressure plasma, especially those operated at low gas temperatures, have become a topic of great interest for the processing of flexible and printed electronic devices due to several benefits such as the reduction of process and reactor costs, the employment of easy-to-handle apparatuses and the easier integration into continuous production lines. In this review, several types of typical atmospheric pressure plasma sources have been addressed, and the processes including surface treatment, texturing and sintering for application to flexible and printed electronic devices have been discussed.

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

Plasma position control in a tokamak reactor around ignition

International Nuclear Information System (INIS)

Carretta, U.; Minardi, E.; Bacelli, N.

1986-01-01

Plasma position control in a tokamak reactor in the phase approaching ignition is closely related to burn control. If ignited burn corresponds to a thermally unstable situation the plasma becomes sensitive to the thermal instability already in the phase when ignition is approached so that the trajectory in the position-pressure (R,p) space becomes effectively unpredictable. For example, schemes involving closed cycles around ignition can be unstable in the heating-cooling phases, and the deviations may be cumulative in time. Reliable plasma control in pressure-position (p, R) space is achieved by beforehand constraining the p, R trajectory rigidly with suitable feedback vertical field stabilization, which is to be established already below ignition. A scheme in which ignition is approached in a stable and automatic way by feedback stabilization on the vertical field is proposed and studied in detail. The values of the gain coefficient ensuring stabilization and the associated p and R excursions are discussed both analytically, with a 0-D approximation including non-linear effects, and numerically with a 1-D code in cylindrical geometry. Profile effects increase the excursions, in particular above ignition. (author)

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

Energy Technology Data Exchange (ETDEWEB)

Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

2011-04-06

Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

Energy Technology Data Exchange (ETDEWEB)

Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

2014-01-15

Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

Atmospheric-pressure plasma technology

International Nuclear Information System (INIS)

Kogelschatz, U

2004-01-01

Major industrial plasma processes operating close to atmospheric pressure are discussed. Applications of thermal plasmas include electric arc furnaces and plasma torches for generation of powders, for spraying refractory materials, for cutting and welding and for destruction of hazardous waste. Other applications include miniature circuit breakers and electrical discharge machining. Non-equilibrium cold plasmas at atmospheric pressure are obtained in corona discharges used in electrostatic precipitators and in dielectric-barrier discharges used for generation of ozone, for pollution control and for surface treatment. More recent applications include UV excimer lamps, mercury-free fluorescent lamps and flat plasma displays

Reactor pressure vessel status report

International Nuclear Information System (INIS)

Strosnider, J.; Wichman, K.; Elliot, B.

1994-12-01

This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

Determination of the neutral oxygen atom density in a plasma reactor loaded with metal samples

Science.gov (United States)

Mozetic, Miran; Cvelbar, Uros

2009-08-01

The density of neutral oxygen atoms was determined during processing of metal samples in a plasma reactor. The reactor was a Pyrex tube with an inner diameter of 11 cm and a length of 30 cm. Plasma was created by an inductively coupled radiofrequency generator operating at a frequency of 27.12 MHz and output power up to 500 W. The O density was measured at the edge of the glass tube with a copper fiber optics catalytic probe. The O atom density in the empty tube depended on pressure and was between 4 and 7 × 1021 m-3. The maximum O density was at a pressure of about 150 Pa, while the dissociation fraction of O2 molecules was maximal at the lowest pressure and decreased with increasing pressure. At about 300 Pa it dropped below 10%. The measurements were repeated in the chamber loaded with different metallic samples. In these cases, the density of oxygen atoms was lower than that in the empty chamber. The results were explained by a drain of O atoms caused by heterogeneous recombination on the samples.

Pressure tube reactors

International Nuclear Information System (INIS)

Natori, Hisahide.

1981-01-01

Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

Plasma properties in a large-volume, cylindrical and asymmetric radio-frequency capacitively coupled industrial-prototype reactor

International Nuclear Information System (INIS)

Lazović, Saša; Puač, Nevena; Spasić, Kosta; Malović, Gordana; Petrović, Zoran Lj; Cvelbar, Uroš; Mozetič, Miran; Radetić, Maja

2013-01-01

We have developed a large-volume low-pressure cylindrical plasma reactor with a size that matches industrial reactors for treatment of textiles. It was shown that it efficiently produces plasmas with only a small increase in power as compared with a similar reactor with 50 times smaller volume. Plasma generated at 13.56 MHz was stable from transition to streamers and capable of long-term continuous operation. An industrial-scale asymmetric cylindrical reactor of simple design and construction enabled good control over a wide range of active plasma species and ion concentrations. Detailed characterization of the discharge was performed using derivative, Langmuir and catalytic probes which enabled determination of the optimal sets of plasma parameters necessary for successful industry implementation and process control. Since neutral atomic oxygen plays a major role in many of the material processing applications, its spatial profile was measured using nickel catalytic probe over a wide range of plasma parameters. The spatial profiles show diffusion profiles with particle production close to the powered electrode and significant wall losses due to surface recombination. Oxygen atom densities range from 10 19 m −3 near the powered electrode to 10 17 m −3 near the wall. The concentrations of ions at the same time are changing from 10 16 to the 10 15 m −3 at the grounded chamber wall. (paper)

Underwater cutting of stainless steel plate and pipe for dismantling reactor pressure vessels

International Nuclear Information System (INIS)

Hamasaki, M.; Tateiwa, F.; Kanatani, F.; Yamashita, S.

1982-01-01

A consumable electrode water jet cutting technique is described. Satisfactory underwater cutting of 80mm stainless steel plate using a current of 2000A and at a water depth of 200mm has been demonstrated. The electrical requirements for this arc welding method applied to cutting were found to be approximately one third those required for conventional plasma arc cutting for the same thickness plate. An application of this technique might be found in the dismantling of atomic reactor pressure vessels, and parts of commercial atomic reactors. (author)

Pressure tube reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Fujino, Michihira.

1980-01-01

Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

Matching of dense plasma focus devices with fission reactors

International Nuclear Information System (INIS)

Harms, A.A.; Heindler, M.

1978-01-01

The potential role of dense plasma focus devices as compact neutron sources for fissile fuel breeding in conjunction with existing fission reactors is considered. It is found that advanced plasma focus devices can be used effectively in conjunction with neutronically efficient fission reactors to constitute ''self-sufficient'' breeders. Correlations among the various parameters such as the power output and conversion ratio of the fission reactor with the neutron yield and capacitor bank energy of the dense plasma focus device are presented and discussed

Applying chemical engineering concepts to non-thermal plasma reactors

Science.gov (United States)

Pedro AFFONSO, NOBREGA; Alain, GAUNAND; Vandad, ROHANI; François, CAUNEAU; Laurent, FULCHERI

2018-06-01

Process scale-up remains a considerable challenge for environmental applications of non-thermal plasmas. Undersanding the impact of reactor hydrodynamics in the performance of the process is a key step to overcome this challenge. In this work, we apply chemical engineering concepts to analyse the impact that different non-thermal plasma reactor configurations and regimes, such as laminar or plug flow, may have on the reactor performance. We do this in the particular context of the removal of pollutants by non-thermal plasmas, for which a simplified model is available. We generalise this model to different reactor configurations and, under certain hypotheses, we show that a reactor in the laminar regime may have a behaviour significantly different from one in the plug flow regime, often assumed in the non-thermal plasma literature. On the other hand, we show that a packed-bed reactor behaves very similarly to one in the plug flow regime. Beyond those results, the reader will find in this work a quick introduction to chemical reaction engineering concepts.

Critical plasma-materials issues for fusion reactor designs

International Nuclear Information System (INIS)

Wilson, K.L.; Bauer, W.

1983-01-01

Plasma-materials interactions are a dominant driving force in the design of fusion power reactors. This paper presents a summary of plasma-materials interactions research. Emphasis is placed on critical aspects related to reactor design. Particular issues to be addressed are plasma edge characterization, hydrogen recycle, impurity introduction, and coating development. Typical wall fluxes in operating magnetically confined devices are summarized. Recent calculations of tritium inventory and first wall permeation, based on laboratory measurements of hydrogen recycling, are given for various reactor operating scenarios. Impurity introduction/wall erosion mechanisms considered include sputtering, chemical erosion, and evaporation (melting). Finally, the advanced material development for in-vessel components is discussed. (author)

Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

Science.gov (United States)

Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

2017-10-01

The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

Results of reactor pressure vessels ISI

International Nuclear Information System (INIS)

Cepcek, S.

1994-01-01

To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

Reactor pressure vessel support

International Nuclear Information System (INIS)

Butti, J.P.

1977-01-01

A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-10-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

Pressurized water reactor with a reactor pressure vessel

International Nuclear Information System (INIS)

Werres, L.

1979-01-01

The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

Novel Composite Hydrogen-Permeable Membranes for Nonthermal Plasma Reactors for the Decomposition of Hydrogen Sulfide

Energy Technology Data Exchange (ETDEWEB)

Morris Argyle; John Ackerman; Suresh Muknahallipatna; Jerry Hamann; Stanislaw Legowski; Gui-Bing Zhao; Sanil John; Ji-Jun Zhang; Linna Wang

2007-09-30

The goal of this experimental project was to design and fabricate a reactor and membrane test cell to dissociate hydrogen sulfide (H{sub 2}S) in a nonthermal plasma and to recover hydrogen (H{sub 2}) through a superpermeable multi-layer membrane. Superpermeability of hydrogen atoms (H) has been reported by some researchers using membranes made of Group V transition metals (niobium, tantalum, vanadium, and their alloys), but it was not achieved at the moderate pressure conditions used in this study. However, H{sub 2}S was successfully decomposed at energy efficiencies higher than any other reports for the high H{sub 2}S concentration and moderate pressures (corresponding to high reactor throughputs) used in this study.

High-beta plasma effects in a low-pressure helicon plasma

International Nuclear Information System (INIS)

Corr, C. S.; Boswell, R. W.

2007-01-01

In this work, high-beta plasma effects are investigated in a low-pressure helicon plasma source attached to a large volume diffusion chamber. When operating above an input power of 900 W and a magnetic field of 30 G a narrow column of bright blue light (due to Ar II radiation) is observed along the axis of the diffusion chamber. With this blue mode, the plasma density is axially very uniform in the diffusion chamber; however, the radial profiles are not, suggesting that a large diamagnetic current might be induced. The diamagnetic behavior of the plasma has been investigated by measuring the temporal evolution of the magnetic field (B z ) and the plasma kinetic pressure when operating in a pulsed discharge mode. It is found that although the electron pressure can exceed the magnetic field pressure by a factor of 2, a complete expulsion of the magnetic field from the plasma interior is not observed. In fact, under our operating conditions with magnetized ions, the maximum diamagnetism observed is ∼2%. It is observed that the magnetic field displays the strongest change at the plasma centre, which corresponds to the maximum in the plasma kinetic pressure. These results suggest that the magnetic field diffuses into the plasma sufficiently quickly that on a long time scale only a slight perturbation of the magnetic field is ever observed

Pressurized water reactor flow arrangement

International Nuclear Information System (INIS)

Gibbons, J.F.; Knapp, R.W.

1980-01-01

A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

Pressurized water reactor systems

International Nuclear Information System (INIS)

Meyer, P.J.

1975-01-01

Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

Plasma-arc reactor for production possibility of powdered nano-size materials

International Nuclear Information System (INIS)

Hadzhiyski, V; Mihovsky, M; Gavrilova, R

2011-01-01

Nano-size materials of various chemical compositions find increasing application in life nowadays due to some of their unique properties. Plasma technologies are widely used in the production of a range of powdered nano-size materials (metals, alloys, oxides, nitrides, carbides, borides, carbonitrides, etc.), that have relatively high melting temperatures. Until recently, the so-called RF-plasma generated in induction plasma torches was most frequently applied. The subject of this paper is the developments of a new type of plasma-arc reactor, operated with transferred arc system for production of disperse nano-size materials. The new characteristics of the PLASMALAB reactor are the method of feeding the charge, plasma arc control and anode design. The disperse charge is fed by a charge feeding system operating on gravity principle through a hollow cathode of an arc plasma torch situated along the axis of a water-cooled wall vertical tubular reactor. The powdered material is brought into the zone of a plasma space generated by the DC rotating transferred plasma arc. The arc is subjected to Auto-Electro-Magnetic Rotation (AEMR) by an inductor serially connected to the anode circuit. The anode is in the form of a water-cooled copper ring. It is mounted concentrically within the cylindrical reactor, with its lower part electrically insulated from it. The electric parameters of the arc in the reactor and the quantity of processed charge are maintained at a level permitting generation of a volumetric plasma discharge. This mode enables one to attain high mean mass temperature while the processed disperse material flows along the reactor axis through the plasma zone where the main physico-chemical processes take place. The product obtained leaves the reactor through the annular anode, from where it enters a cooling chamber for fixing the produced nano-structure. Experiments for AlN synthesis from aluminium power and nitrogen were carried out using the plasma reactor

Pressurized water reactor inspection procedures

International Nuclear Information System (INIS)

Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

1998-01-01

Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

Leak detector for reactor pressure vessel

International Nuclear Information System (INIS)

Morimoto, Mikio.

1991-01-01

A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

N Reactor pressure tube 2566 postirradiation examination

International Nuclear Information System (INIS)

Scott, K.V.

1978-01-01

Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

Plasma engineering analysis of a small torsatron reactor

International Nuclear Information System (INIS)

Lacatski, J.T.; Houlberg, W.A.; Uckan, N.A.

1985-10-01

This study examines the plasma physics and reactor engineering feasibility of a small, medium aspect ratio, high-beta, l = 2, D-T torsatron power reactor, based on the magnetic configuration of the Advanced Toroidal Facility, Oak Ridge National Laboratory. Plasma analyses are performed to assess whether confinement in a small, average radius plasma is sufficient to yield an ignited or high-Q driven device. Much of the physics assessment focuses on an evaluation of the radial electric field created by the nonambipolar particle flux. Detailed transport simulations are done with both fixed and self-consistent evolution of the radial electric field. Basic reactor engineering considerations taken into account are neutron wall loading, maximum magnetic field at the helical coils, coil shield thickness, and tritium breeding blanket-shield thickness

Plasma core reactor applications

International Nuclear Information System (INIS)

Latham, T.S.; Rodgers, R.J.

1976-01-01

Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

Ultrasound periodic inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Haniger, L.

1980-01-01

Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

FABRICATION OF CNTS BY TOLUENE DECOMPOSITION IN A NEW REACTOR BASED ON AN ATMOSPHERIC PRESSURE PLASMA JET COUPLED TO A CVD SYSTEM

Directory of Open Access Journals (Sweden)

FELIPE RAMÍREZ-HERNÁNDEZ

2017-03-01

Full Text Available Here, we present a method to produce carbon nanotubes (CNTs based on the coupling between two conventional techniques used for the preparation of nanostructures: an arc-jet as a source of plasma and a chemical vapour deposition (CVD system. We call this system as an “atmospheric pressure plasma (APP-enhanced CVD” (APPE-CVD. This reactor was used to grow CNTs on non-flat aluminosilicate substrates by the decomposition of toluene (carbon source in the presence of ferrocene (as a catalyst. Both, CNTs and by-products of carbon were collected at three different temperatures (780, 820 and 860 °C in different regions of the APPE-CVD system. These samples were analysed by thermogravimetric analysis (TGA and DTG, scanning electron microscopy (SEM and Raman spectroscopy in order to determine the effect of APP on the thermal stability of the as-grown CNTs. It was found that the amount of metal catalyst in the synthesised CNTs is reduced by applying APP, being 820 °C the optimal temperature to produce CNTs with a high yield and carbon purity (95 wt. %. In contrast, when the synthesis temperature was fixed at 780 °C or 860 °C, amorphous carbon or CNTs with different structural defects, respectively, was formed through APEE-CVD reactor. We recommended the use of non-flat aluminosilicate particles as supports to increase CNT yield and facilitate the removal of deposits from the substrate surface. The approach that we implemented (to synthesise CNTs by using the APPE-CVD reactor may be useful to produce these nanostructures on a gram-scale for use in basic studies. The approach may also be scaled up for mass production.

Reactor Pressure Vessel (RPV) Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

Comparative Analysis of Carbon Plasma in Arc and RF Reactors

International Nuclear Information System (INIS)

Todorovic-Markovic, B.; Markovic, Z.; Mohai, I.; Szepvolgyi, J.

2004-01-01

Results on studies of molecular spectra emitted in the initial stages of fullerene formation during the processing of graphite powder in induction RF reactor and evaporation of graphite electrodes in arc reactor are presented in this paper. It was found that C2 radicals were dominant molecular species in both plasmas. C2 radicals have an important role in the process of fullerene synthesis. The rotational-vibrational temperatures of C2 and CN species were calculated by fitting the experimental spectra to the simulated ones. The results of optical emission study of C2 radicals generated in carbon arc plasma have shown that rotational temperature of C2 species depends on carbon concentration and current intensity significantly. The optical emission study of induction RF plasma and SEM analysis of graphite powder before and after plasma treatment have shown that evaporation of the processed graphite powder depends on feed rate and composition of gas phase significantly. Based on the obtained results, it was concluded that in the plasma region CN radicals could be formed by the reaction of C2 species with atomic nitrogen at smaller loads. At larger feed rate of graphite powder, CN species were produced by surface reaction of the hot carbon particles with nitrogen atoms. The presence of nitrogen in induction RF plasma reduces the fullerene yield significantly. The fullerene yield obtained in two different reactors was: 13% in arc reactor and 4.1% in induction RF reactor. However, the fullerene production rate was higher in induction RF reactor-6.4 g/h versus 1.7 g/h in arc reactor

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

Energy Technology Data Exchange (ETDEWEB)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-04-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

International Nuclear Information System (INIS)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-01-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

Plasma spheroidization of nickel powders in a plasma reactor

Indian Academy of Sciences (India)

Unknown

and size of the particles are among critical parameters ... shape components, which helps to conserve scarce raw materials. There are several methods of producing rapidly solidi- ... spherical nickel powders using the d.c. plasma reactor is.

Desizing of Starch Containing Cotton Fabrics Using Near Atmospheric Pressure, Cold DC Plasma Treatment

Science.gov (United States)

Prasath, A.; Sivaram, S. S.; Vijay Anand, V. D.; Dhandapani, Saravanan

2013-03-01

An attempt has been made to desize the starch containing grey cotton fabrics using the DC plasma with oxygen as the gaseous medium. Process conditions of the plasma reactor were optimized in terms of distance between the plates (3.2 cm), applied voltage (600 V) and applied pressure (0.01 bar) to obtain maximum desizing efficiency. No discolouration was observed in the hot water extracts of the desized sample in presence of iodine though relatively higher solvent extractable impurities (4.53 %) were observed in the plasma desized samples compared to acid desized samples (3.38 %). Also, significant weight loss, improvements in plasma desized samples were observed than that of grey fabrics in terms of drop absorbency.

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

Atomic hydrogen determination in medium-pressure microwave discharge hydrogen plasmas via emission actinometry

International Nuclear Information System (INIS)

Geng Zicai; Xu Yong; Yang Xuefeng; Wang Weiguo; Zhu Aimin

2005-01-01

Atomic hydrogen plays an important role in the chemical vapour deposition of functional materials, plasma etching and new approaches to the chemical synthesis of hydrogen-containing compounds. This work reports experimental determinations of atomic hydrogen in microwave discharge hydrogen plasmas formed from the TM 01 microwave mode in an ASTeX-type reactor, via optical emission spectroscopy using Ar as an actinometer. The relative intensities of the H atom Balmer lines and Ar-750.4 nm emissions as functions of input power and gas pressure have been investigated. At an input microwave power density of 13.5 W cm -3 , the approximate hydrogen dissociation fractions calculated from electron-impact excitation and quenching cross sections in the literature, decreased from ∼0.08 to ∼0.03 as the gas pressure was increased from 5 to 25 Torr. The influences of the above cross sections, and the electron and gas temperatures of the plasmas on the determination of the hydrogen dissociation fraction data have been discussed

Fuel assembly for pressure loss variable PWR type reactor

International Nuclear Information System (INIS)

Yoshikuni, Masaaki.

1993-01-01

In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

Large area atmospheric-pressure plasma jet

Science.gov (United States)

Selwyn, Gary S.; Henins, Ivars; Babayan, Steve E.; Hicks, Robert F.

2001-01-01

Large area atmospheric-pressure plasma jet. A plasma discharge that can be operated at atmospheric pressure and near room temperature using 13.56 MHz rf power is described. Unlike plasma torches, the discharge produces a gas-phase effluent no hotter than 250.degree. C. at an applied power of about 300 W, and shows distinct non-thermal characteristics. In the simplest design, two planar, parallel electrodes are employed to generate a plasma in the volume therebetween. A "jet" of long-lived metastable and reactive species that are capable of rapidly cleaning or etching metals and other materials is generated which extends up to 8 in. beyond the open end of the electrodes. Films and coatings may also be removed by these species. Arcing is prevented in the apparatus by using gas mixtures containing He, which limits ionization, by using high flow velocities, and by properly spacing the rf-powered electrode. Because of the atmospheric pressure operation, there is a negligible density of ions surviving for a sufficiently long distance beyond the active plasma discharge to bombard a workpiece, unlike the situation for low-pressure plasma sources and conventional plasma processing methods.

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Combustion flame-plasma hybrid reactor systems, and chemical reactant sources

Science.gov (United States)

Kong, Peter C

2013-11-26

Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

Atmospheric pressure cold plasma treatment of cellulose based fillers for wood plastic composites

Science.gov (United States)

Lekobou, William; Englund, Karl; Pedrow, Patrick; Scudiero, Louis

2011-10-01

The main challenge of wood plastic composites (WPC) resides in the low interfacial adhesion due to incompatibility between the cellulose based filler that has a polar surface and most common matrixes, polyolefins which are non-polar. Plasma treatment is a promising technique for surface modification and its implementation into the processing of WPC would provide this industry with a versatile and nearly environmentally benign manufacturing tool. Our investigation aims at designing a cold atmospheric pressure plasma reactor for coating fillers with a hydrophobic material prior to compounding with the matrix. Deposition was achieved with our reactor that includes an array of high voltage needles, a grounded metal mesh, Ar as carrier gas and C2H2 as the precursor molecule. Parameters studied have included gas feed rates and applied voltage; FTIR, ESCA, AFM and SEM imaging were used for film diagnostics. We will also report on deposition rate and its dependence on radial and axial position as well as the effects of plasma-polymerized acetylene on the surface free energy of cellulose based substrates.

Atmospheric pressure microwave plasma system with ring waveguide

International Nuclear Information System (INIS)

Liu Liang; Zhang Guixin; Zhu Zhijie; Luo Chengmu

2007-01-01

Some scientists used waveguide as the cavity to produce a plasma jet, while large volume microwave plasma was relatively hard to get in atmospheric pressure. However, a few research institutes have already developed devices to generate large volume of atmospheric pressure microwave plasma, such as CYRANNUS and SLAN series, which can be widely applied. In this paper, present a microwave plasma system with ring waveguide to excite large volume of atmospheric pressure microwave plasma, plot curves on theoretical disruption electric field of some working gases, emulate the cavity through software, measure the power density to validate and show the appearance of microwave plasma. At present, large volume of argon and helium plasma have already been generated steadily by atmospheric pressure microwave plasma system. This research can build a theoretical basis of microwave plasma excitation under atmospheric pressure and will be useful in study of the device. (authors)

Effect of ambient pressure on the crystalline phase of nano TiO2 particles synthesized by a dc thermal plasma reactor

International Nuclear Information System (INIS)

Banerjee, I.; Karmakar, Soumen; Kulkarni, Naveen V.; Nawale, Ashok B.; Mathe, V. L.; Das, A. K.; Bhoraskar, S. V.

2010-01-01

The synthesis of nanoparticles of titanium dioxide (TiO 2 ) with varying percentages of anatase and rutile phases is reported. This was achieved by controlling the operating pressure in a transferred-arc, direct current thermal plasma reactor in which titanium vapors are evaporated, and then exposed to ambient oxygen. The average particle size remained around 15 nm in each case. The crystalline structure of the as-synthesized nanoparticles of TiO 2 was studied with X-ray diffraction analysis; whereas the particle morphology was investigated with the help of transmission electron microscopy. The precursor species responsible for the growth of these nanoparticles was studied with the help of optical emission spectroscopy. As inferred from the X-ray diffraction analysis, the relative abundance of anatase TiO 2 was found to be dominant when synthesized at 760 Torr, and the same showed a decreasing trend with decreasing chamber pressure. The study also reveals that anatase TiO 2 is a more effective photocatalytic agent in degrading methylene blue by comparison to its rutile phase.

Prestressed concrete pressure vessels for nuclear reactors - 1973

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

International Nuclear Information System (INIS)

Sun Yunxue; Shao Jin

2015-01-01

Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

SCW Pressure-Channel Nuclear Reactor Some Design Features

Science.gov (United States)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Plasma Diagnostics in High Density Reactors

International Nuclear Information System (INIS)

Daltrini, A. M.; Moshkalyov, S.; Monteiro, M. J. R.; Machida, M.; Kostryukov, A.; Besseler, E.; Biasotto, C.; Diniz, J. A.

2006-01-01

Langmuir electric probes and optical emission spectroscopy diagnostics were developed for applications in high density plasmas. These diagnostics were employed in two plasma sources: an electron cyclotron resonance (ECR) plasma and an RF driven inductively coupled plasma (ICP) plasma. Langmuir probes were tested using a number of probing dimensions, probe tip materials, circuits for probe bias and filters. Then, the results were compared with the optical spectroscopy measurements. With these diagnostics, analyses of various plasma processes were performed in both reactors. For example, it has been shown that species like NH radicals generated in gas phase can have critical impact on films deposited by ECR plasmas. In the ICP source, plasmas in atomic and molecular gases were shown to have different spatial distributions, likely due to nonlocal electron heating. The low-to-high density transitions in the ICP plasma were also studied. The role of metastables is shown to be significant in Ar plasmas, in contrast to plasmas with additions of molecular gases

Numerical modeling of disperse material evaporation in axisymmetric thermal plasma reactor

Directory of Open Access Journals (Sweden)

Stefanović Predrag Lj.

2003-01-01

Full Text Available A numerical 3D Euler-Lagrangian stochastic-deterministic (LSD model of two-phase flow laden with solid particles was developed. The model includes the relevant physical effects, namely phase interaction, panicle dispersion by turbulence, lift forces, particle-particle collisions, particle-wall collisions, heat and mass transfer between phases, melting and evaporation of particles, vapour diffusion in the gas flow. It was applied to simulate the processes in thermal plasma reactors, designed for the production of the ceramic powders. Paper presents results of extensive numerical simulation provided (a to determine critical mechanism of interphase heat and mass transfer in plasma flows, (b to show relative influence of some plasma reactor parameters on solid precursor evaporation efficiency: 1 - inlet plasma temperature, 2 - inlet plasma velocity, 3 - particle initial diameter, 4 - particle injection angle a, and 5 - reactor wall temperature, (c to analyze the possibilities for high evaporation efficiency of different starting solid precursors (Si, Al, Ti, and B2O3 powder, and (d to compare different plasma reactor configurations in conjunction with disperse material evaporation efficiency.

EPR (European Pressurized Reactor)

International Nuclear Information System (INIS)

2015-01-01

This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

International Nuclear Information System (INIS)

Bros, J.; Ballesteros, A.; Lopez, A.

1993-01-01

Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

Standard Technical Specifications for Westinghouse pressurized water reactors

International Nuclear Information System (INIS)

Virgilio, M.J.

1980-09-01

The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Pressurizer pump reliability analysis high flux isotope reactor

International Nuclear Information System (INIS)

Merryman, L.; Christie, B.

1993-01-01

During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

Degradation of aqueous phenol solutions by coaxial DBD reactor

Science.gov (United States)

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).

Diagnostics of atmospheric pressure air plasmas

International Nuclear Information System (INIS)

Laux, C.O.; Kruger, C.H.; Zare, R.N.

2001-01-01

Atmospheric pressure air plasmas are often thought to be in Local Thermodynamics Equilibrium (LTE) owing to fast interspecies collisional exchanges at high pressure. As will be seen here, this assumption cannot be relied upon, particularly with respect to optical diagnostics. Large velocity gradients in flowing plasmas and/or elevated electron temperatures created by electrical discharges can result in large departures from chemical and thermal equilibrium. Diagnostic techniques based on optical emission spectroscopy (OES) and Cavity Ring-Down Spectroscopy (CRDS) have been developed and applied at Stanford University to the investigation of atmospheric pressure plasmas under conditions ranging from thermal and chemical equilibrium to thermochemical nonequilibrium. This article presents a review of selected temperature and species concentration measurement techniques useful for the study of air and nitrogen plasmas

Synthesis of Carbon Nanotubes in Thermal Plasma Reactor at Atmospheric Pressure

Directory of Open Access Journals (Sweden)

Lukasz Szymanski

2017-02-01

Full Text Available 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.

Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

International Nuclear Information System (INIS)

Chou, Hsoung-Wei; Huang, Chin-Cheng

2016-01-01

Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

Thermal annealing of an embrittled reactor pressure vessel

International Nuclear Information System (INIS)

Mager, T.R.; Dragunov, Y.G.; Leitz, C.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

Limit regulation system for pressurized water nuclear reactors

International Nuclear Information System (INIS)

Aleite, W.; Bock, H.W.

1976-01-01

Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

International Nuclear Information System (INIS)

Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

2001-01-01

The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

CFD simulation analysis and validation for CPR1000 pressurized water reactor

International Nuclear Information System (INIS)

Zhang Mingqian; Ran Xiaobing; Liu Yanwu; Yu Xiaolei; Zhu Mingli

2013-01-01

Background: With the rapid growth in the non-nuclear area for industrial use of Computational fluid dynamics (CFD) which has been accompanied by dramatically enhanced computing power, the application of CFD methods to problems relating to Nuclear Reactor Safety (NRS) is rapidly accelerating. Existing research data have shown that CFD methods could predict accurately the pressure field and the flow repartition in reactor lower plenum. But simulations for the full domain of the reactor have not been reported so far. Purpose: The aim is to determine the capabilities of the codes to model accurately the physical phenomena which occur in the full reactor vessel. Methods: The flow field of the CPR1000 reactor which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX. The pressure loss in reactor pressure vessel, the hydraulic loads of guide tubes and support columns, and the bypass flow of head dome were obtained by calculations for the full domain of the reactor. The results were validated by comparing with the determined reference value of the operating nuclear plant (LingAo nuclear plant), and the transient simulation was conducted in order to better understand the flow in reactor pressure vessel. Results: It was shown that the predicted pressure loss with CFD code was slightly different with the determined value (10% relative deviation for the total pressure loss), the hydraulic loads were less than the determined value with maximum relative deviation 50%, and bypass flow of head dome was approximately the same with determined value. Conclusion: This analysis practice predicts accurately the physical phenomena which occur in the full reactor vessel, and can be taken as a guidance for the nuclear plant design development and improve our understanding of reactor flow phenomena. (authors)

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

International Nuclear Information System (INIS)

Ware, A.G.

1988-01-01

This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

Note: Interpolation for evaluation of a two-dimensional spatial profile of plasma densities at low gas pressures

International Nuclear Information System (INIS)

Oh, Se-Jin; Kim, Young-Chul; Chung, Chin-Wook

2011-01-01

An interpolation algorithm for the evaluation of the spatial profile of plasma densities in a cylindrical reactor was developed for low gas pressures. The algorithm is based on a collisionless two-dimensional fluid model. Contrary to the collisional case, i.e., diffusion fluid model, the fitting algorithm depends on the aspect ratio of the cylindrical reactor. The spatial density profile of the collisionless fitting algorithm is presented in two-dimensional images and compared with the results of the diffusion fluid model.

Foundations of atmospheric pressure non-equilibrium plasmas

Science.gov (United States)

Bruggeman, Peter J.; Iza, Felipe; Brandenburg, Ronny

2017-12-01

Non-equilibrium plasmas have been intensively studied over the past century in the context of material processing, environmental remediation, ozone generation, excimer lamps and plasma display panels. Research on atmospheric pressure non-equilibrium plasmas intensified over the last two decades leading to a large variety of plasma sources that have been developed for an extended application range including chemical conversion, medicine, chemical analysis and disinfection. The fundamental understanding of these discharges is emerging but there remain a lot of unexplained phenomena in these intrinsically complex plasmas. The properties of non-equilibrium plasmas at atmospheric pressure span over a huge range of electron densities as well as heavy particle and electron temperatures. This paper provides an overview of the key underlying processes that are important for the generation and stabilization of atmospheric pressure non-equilibrium plasmas. The unique physical and chemical properties of theses discharges are also summarized.

Pressure suppression facility for reactor container

International Nuclear Information System (INIS)

Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

1993-01-01

In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

Functional systems of a pressurized water reactor

International Nuclear Information System (INIS)

Heinzel, V.

1982-01-01

The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

Pressurized water reactors: the EPR project

International Nuclear Information System (INIS)

Py, J.P.; Yvon, M.

2007-01-01

EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

Heat insulation device for reactor pressure vessel in water

International Nuclear Information System (INIS)

Nakamura, Heiichiro; Tanaka, Yoshimi.

1993-01-01

Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

Excess-pressure suppression device in a reactor container

International Nuclear Information System (INIS)

Nishio, Masahide

1985-01-01

Purpose: To reliably decrease the radioactivity of radioactive gases when they are released externally. Constitution: The exit of a gas exhaust pipe for discharging gases in a reactor container, on generation of an excess pressure in the reactor container upon loss of coolant accident, is adapted to be always fluided in the cooling tank. Then, the exhaust gases discharged in the cooling tank is realeased to the atmosphere. In this way, the excess pressure in the reactor container can be prevented previously and the radioactivity of the gases released externally is significantly reduced by the scrubbing effect. (Kamimura, M.)

The future 700 MWe pressurized heavy water reactor

International Nuclear Information System (INIS)

Bhardwaj, S.A.

2006-01-01

The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

Effects of pressure anisotropy on plasma transport

International Nuclear Information System (INIS)

Zawaideh, E.; Najmabadi, F.; Conn, R.W.

1986-03-01

In a recent paper a new set of generalized two-field equations is derived which describes plasma transport along the field lines of a space and time dependent magnetic field. These equations are valid for collisional to weakly collisional plasmas; they reduce to the conventional fluid equations of Braginskii for highly collisional plasmas. An important feature of these equations is that the anisotropy in the ion pressure is explicitly included. In this paper, these generalized transport equations are applied to a model problem of plasma flow through a magnetic mirror field. The profiles of the plasma parameters (density, flow speed, and pressures) are numerically calculated for plasma in different collisionality regimes. These profiles are explained by examining the competing terms in the transport equation. The pressure anisotropy is found to profoundly impact the plasma flow behavior. As a result, the new generalized equations predict flow behavior more accurately than the conventional transport equations. A large density and pressure drop is predicted as the flow passes through a magnetic mirror. Further, the new equations uniquely predict oscillations in the density profile, an effect missing in results from the conventional equations

How to replace a reactor pressure vessel

International Nuclear Information System (INIS)

Huber, R.

1996-01-01

A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

Tailoring nanomaterial products through electrode material and oxygen partial pressure in a mini-arc plasma reactor

International Nuclear Information System (INIS)

Cui Shumao; Mattson, Eric C.; Lu, Ganhua; Hirschmugl, Carol; Gajdardziska-Josifovska, Marija; Chen Junhong

2012-01-01

Nanomaterials with controllable morphology and composition are synthesized by a simple one-step vapor condensation process using a mini-arc plasma source. Through systematic investigation of mini-arc reactor parameters, the roles of carrier gas, electrode material, and precursor on producing diverse nanomaterial products are revealed. Desired nanomaterial products, including tungsten oxide nanoparticles (NPs), tungsten oxide nanorods (NRs), tungsten oxide and tin oxide NP mixtures and pure tin dioxide NPs can thus be obtained by tailoring reaction conditions. The amount of oxygen in the reactor is critical to determining the final nanomaterial product. Without any precursor material present, a lower level of oxygen in the reactor favors the production of W 18 O 49 NRs with tungsten as cathode, while a high level of oxygen produces more round WO 3 NPs. With the presence of a precursor material, amorphous particles are favored with a high ratio of argon:oxygen. Oxygen is also found to affect tin oxide crystallization from its amorphous phase in the thermal annealing. Results from this study can be used for guiding gas phase nanomaterial synthesis in the future.

On atmospheric-pressure non-equilibrium plasma jets and plasma bullets

International Nuclear Information System (INIS)

Lu, X; Laroussi, M; Puech, V

2012-01-01

Atmospheric-pressure non-equilibrium plasma jets (APNP-Js), which generate plasma in open space rather than in a confined discharge gap, have recently been a topic of great interest. In this paper, the development of APNP-Js will be reviewed. Firstly, the APNP-Js are grouped based on the type of gas used to ignite them and their characteristics are discussed in detail. Secondly, one of the most interesting phenomena of APNP-Js, the ‘plasma bullet’, is discussed and its behavior described. Thirdly, the very recent developments on the behavior of plasma jets when launched in a controlled environment and pressure are also introduced. This is followed by a discussion on the interaction between plasma jets. Finally, perspectives on APNP-J research are presented. (paper)

Theory and measurements of electrophoretic effects in monolith, fixed-bed, and fluidized-bed plasma reactors

International Nuclear Information System (INIS)

Morin, T.J.

1989-01-01

Pressure gradients and secondary flow fields generated by the passage of electrical current in a d.c. gas discharge or gas laser are topics of longstanding interest in the gaseous electronics literature. These hydrodynamic effects of space charge fields and charged particle density gradients have been principally exploited in the development of gas separation and purification processes. In recent characterization studies of fixed-bed and fluidized-bed plasma reactors several anomalous flow features have been observed. These reactors involve the contacting of a high-frequency, resonantly-sustained, disperse gas discharge with granular solids in a fixed or fluidized bed. Anomalies in the measured pressure drops and fluidization velocities have motivated the development of an appropriate theoretical approach to, and some additional experimental investigations of electrophoretic effects in disperse gas discharges. In this paper, a theory which includes the effects of space charge and diffusion is used to estimate the electric field and charged particle density profiles. These profiles are then used to calculate velocity fields and gas flow rates for monolith, fixed-bed, and fluidized-bed reactors. These results are used to rationalize measurements of gas flow rates and axial pressure gradients in high-frequency disperse gas discharges with and without an additional d.c. axial electric field

Integrity of Magnox reactor steel pressure vessels

International Nuclear Information System (INIS)

Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

1992-01-01

The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Numerical Modelling of Wood Gasification in Thermal Plasma Reactor

Czech Academy of Sciences Publication Activity Database

Hirka, Ivan; Živný, Oldřich; Hrabovský, Milan

2017-01-01

Roč. 37, č. 4 (2017), s. 947-965 ISSN 0272-4324 Institutional support: RVO:61389021 Keywords : Plasma modelling * CFD * Thermal plasma reactor * Biomass * Gasification * Syngas Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.355, year: 2016 https://link.springer.com/article/10.1007/s11090-017-9812-z

Low pressure plasma discharges for the sterilization and decontamination of surfaces

International Nuclear Information System (INIS)

Rossi, F; Rauscher, H; Hasiwa, M; Gilliland, D; Kylian, O

2009-01-01

The mechanisms of sterilization and decontamination of surfaces are compared in direct and post discharge plasma treatments in two low-pressure reactors, microwave and inductively coupled plasma. It is shown that the removal of various biomolecules, such as proteins, pyrogens or peptides, can be obtained at high rates and low temperatures in the inductively coupled plasma (ICP) by using Ar/O 2 mixtures. Similar efficiency is obtained for bacterial spores. Analysis of the discharge conditions illustrates the role of ion bombardment associated with O radicals, leading to a fast etching of organic matter. By contrast, the conditions obtained in the post discharge lead to much lower etching rates but also to a chemical modification of pyrogens, leading to their de-activation. The advantages of the two processes are discussed for the application to the practical case of decontamination of medical devices and reduction of hospital infections, illustrating the advantages and drawbacks of the two approaches.

Low pressure plasma discharges for the sterilization and decontamination of surfaces

Energy Technology Data Exchange (ETDEWEB)

Rossi, F; Rauscher, H; Hasiwa, M; Gilliland, D [European Commission, Joint Research Centre, Institute for Health and Consumer Protection, Via E. Fermi 2749, 21027 Ispra (Vatican City State, Holy See) (Italy); Kylian, O [Faculty of Mathematics and Physics, Charles University, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)], E-mail: francois.rossi@jrc.ec.europa.eu

2009-11-15

The mechanisms of sterilization and decontamination of surfaces are compared in direct and post discharge plasma treatments in two low-pressure reactors, microwave and inductively coupled plasma. It is shown that the removal of various biomolecules, such as proteins, pyrogens or peptides, can be obtained at high rates and low temperatures in the inductively coupled plasma (ICP) by using Ar/O{sub 2} mixtures. Similar efficiency is obtained for bacterial spores. Analysis of the discharge conditions illustrates the role of ion bombardment associated with O radicals, leading to a fast etching of organic matter. By contrast, the conditions obtained in the post discharge lead to much lower etching rates but also to a chemical modification of pyrogens, leading to their de-activation. The advantages of the two processes are discussed for the application to the practical case of decontamination of medical devices and reduction of hospital infections, illustrating the advantages and drawbacks of the two approaches.

Pressure balance between lobe and plasma sheet

International Nuclear Information System (INIS)

Baumjohann, W.; Paschmann, G.; Luehr, H.

1990-01-01

Using eight months of AMPTE/IRM plasma and magnetic field data, the authors have done a statistical survey on the balance of total (thermal and magnetic) pressure in the Earth's plasma sheet and tail lobe. About 300,000 measurements obtained in the plasma sheet and the lobe were compared for different levels of magnetic activity as well as different distances from the Earth. The data show that lobe and plasma sheet pressure balance very well. Even in the worst case they do not deviate by more than half of the variance in the data itself. Approximately constant total pressure was also seen during a quiet time pass when IRM traversed nearly the whole magnetotail in the vertical direction, from the southern hemisphere lobe through the neutral sheet and into the northern plasma sheet boundary layer

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

Investigation of the failure of a reactor pressure vessel by plastic instability

International Nuclear Information System (INIS)

Laemmer, H.; Ritter, B.

1994-01-01

A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

Plasma startup patterns in tokamak reactors

International Nuclear Information System (INIS)

Maki, Koichi; Tone, Tatsuzo.

1983-01-01

Plasma startup patterns are studied from the viewpoint of net power loss represented by the total power loss less the α-particle heating power. The existence is shown of a critical temperature of plasma at which the net power loss becomes independent of plasma density. Observations are made which indicate that the net power loss decreases with lowering plasma density in the range below the critical temperature and vice versa, whether governed by empirical or trapped-ion scaling laws. A startup pattern is presented which minimizes the net power loss during startup, and which prescribes that: (1) The plasma density should be kept as low as possible until the plasma is heated up to the critical temperature; (2) thereafter, the plasma density should be increased to its steady state value while retaining the critical temperature; and (3) finally, with the density kept constant, the temperature should be further raised to its steady state value. The net power loss at critical temperature represents the lower limit of heating power required to bring the plasma to steady state in tokamak reactors. (author)

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

Future view of total energy system and reactor engineering and reactor physics

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

Thermohydraulic analysis of pressurized water reactors

International Nuclear Information System (INIS)

Veloso, M.A.

1980-01-01

The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

International Nuclear Information System (INIS)

Vito, D.J.

1980-12-01

The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

Pressurized water reactor fuel rod design methodology

International Nuclear Information System (INIS)

Silva, A.T.; Esteves, A.M.

1988-08-01

The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

Light Water Reactor-Pressure Vessel Surveillance project computer system

International Nuclear Information System (INIS)

Merriman, S.H.

1980-10-01

A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion

Directory of Open Access Journals (Sweden)

Istadi Istadi

2007-10-01

Full Text Available A catalytic - DBD plasma reactor was designed and developed for co-generation of synthesis gas and C2+ hydrocarbons from methane. A hybrid Artificial Neural Network - Genetic Algorithm (ANN-GA was developed to model, simulate and optimize the reactor. Effects of CH4/CO2 feed ratio, total feed flow rate, discharge voltage and reactor wall temperature on the performance of catalytic DBD plasma reactor was explored. The Pareto optimal solutions and corresponding optimal operating parameters ranges based on multi-objectives can be suggested for catalytic DBD plasma reactor owing to two cases, i.e. simultaneous maximization of CH4 conversion and C2+ selectivity, and H2 selectivity and H2/CO ratio. It can be concluded that the hybrid catalytic DBD plasma reactor is potential for co-generation of synthesis gas and higher hydrocarbons from methane and carbon dioxide and showed better than the conventional fixed bed reactor with respect to CH4 conversion, C2+ yield and H2 selectivity for CO2 OCM process. © 2007 BCREC UNDIP. All rights reserved.[Presented at Symposium and Congress of MKICS 2007, 18-19 April 2007, Semarang, Indonesia][How to Cite: I. Istadi, N.A.S. Amin. (2007. Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion. Bulletin of Chemical Reaction Engineering and Catalysis, 2 (2-3: 37-44.  doi:10.9767/bcrec.2.2-3.8.37-44][How to Link/DOI: http://dx.doi.org/10.9767/bcrec.2.2-3.8.37-44 || or local: http://ejournal.undip.ac.id/index.php/bcrec/article/view/8][Cited by: Scopus 1 |

Observation of intense beam in low pressure from IPR Plasma Focus facility

International Nuclear Information System (INIS)

Kumar, R.; Shyam, A.; Chaturvedi, S.; Lathi, D.; Sarkar, Partha; Chaudhari, V.; Verma, R.; Shukla, R.; Debnath, K.; Sonara, J.; Shah, K.; Adhikary, B.

2004-01-01

Full text: Plasma focus (PF) is a powerful source of various ionizing radiation such as charged particles beam (ions and electrons), X-ray, neutrons etc. This device can operate from energy level of 50J to 1MJ. Plasma Focus is relatively small, simple and cheap in comparison with other radiation sources based on isotopes, accelerators and fusion reactors. Radiation pulse from PF is strong and very short. Now with the new pulsed power technology this device can be operated repeatedly with enhanced lifetime. All these features make plasma focus a versatile device for academic as well as industrial interest such as hot plasma physics and plasma collective processes, equation of state of matter under extreme conditions, material science including material characterization, dynamic equation control, and surface modification and destruction test. Intense burst of neutrons have been observed from a low energy (3.6 kJ) Mather type plasma focus device operated in 0.4 Torr pressure of deuterium medium at IPR. The emitted neutrons (10 9 /shot), that are accompanied by a strong hard X-ray pulse, were found to be having energy up to 3.26 MeV in the axial direction of the device

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

1975-01-01

The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

Atmospheric-Pressure Plasma Interaction with Soft Materials as Fundamental Processes in Plasma Medicine.

Science.gov (United States)

Takenaka, Kosuke; Miyazaki, Atsushi; Uchida, Giichiro; Setsuhara, Yuichi

2015-03-01

Molecular-structure variation of organic materials irradiated with atmospheric pressure He plasma jet have been investigated. Optical emission spectrum in the atmospheric-pressure He plasma jet has been measured. The spectrum shows considerable emissions of He lines, and the emission of O and N radicals attributed to air. Variation in molecular structure of Polyethylene terephthalate (PET) film surface irradiated with the atmospheric-pressure He plasma jet has been observed via X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FT-IR). These results via XPS and FT-IR indicate that the PET surface irradiated with the atmospheric-pressure He plasma jet was oxidized by chemical and/or physical effect due to irradiation of active species.

Pressurized water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

EPR (European Pressurized water Reactor) The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-01-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

Fourier transform infrared absorption spectroscopy characterization of gaseous atmospheric pressure plasmas with 2 mm spatial resolution

Energy Technology Data Exchange (ETDEWEB)

Laroche, G. [Laboratoire d' Ingenierie de Surface, Centre de Recherche sur les Materiaux Avances, Departement de genie des mines, de la metallurgie et des materiaux, Universite Laval, 1065, avenue de la Medecine, Quebec G1V 0A6 (Canada); Centre de recherche du CHUQ, Hopital St Francois d' Assise, 10, rue de l' Espinay, local E0-165, Quebec G1L 3L5 (Canada); Vallade, J. [Laboratoire Procedes, Materiaux et Energie Solaire, PROMES, CNRS, Technosud, Rambla de la Thermodynamique, F-66100 Perpignan (France); Agence de l' environnement et de la Ma Latin-Small-Letter-Dotless-I -carettrise de l' Energie, 20, avenue du Gresille, BP 90406, F-49004 Angers Cedex 01 (France); Bazinette, R.; Hernandez, E.; Hernandez, G.; Massines, F. [Laboratoire Procedes, Materiaux et Energie Solaire, PROMES, CNRS, Technosud, Rambla de la Thermodynamique, F-66100 Perpignan (France); Nijnatten, P. van [OMT Solutions bv, High Tech Campus 9, 5656AE Eindhoven (Netherlands)

2012-10-15

This paper describes an optical setup built to record Fourier transform infrared (FTIR) absorption spectra in an atmospheric pressure plasma with a spatial resolution of 2 mm. The overall system consisted of three basic parts: (1) optical components located within the FTIR sample compartment, making it possible to define the size of the infrared beam (2 mm Multiplication-Sign 2 mm over a path length of 50 mm) imaged at the site of the plasma by (2) an optical interface positioned between the spectrometer and the plasma reactor. Once through the plasma region, (3) a retro-reflector module, located behind the plasma reactor, redirected the infrared beam coincident to the incident path up to a 45 Degree-Sign beamsplitter to reflect the beam toward a narrow-band mercury-cadmium-telluride detector. The antireflective plasma-coating experiments performed with ammonia and silane demonstrated that it was possible to quantify 42 and 2 ppm of these species in argon, respectively. In the case of ammonia, this was approximately three times less than this gas concentration typically used in plasma coating experiments while the silane limit of quantification was 35 times lower. Moreover, 70% of the incoming infrared radiation was focused within a 2 mm width at the site of the plasma, in reasonable agreement with the expected spatial resolution. The possibility of reaching this spatial resolution thus enabled us to measure the gaseous precursor consumption as a function of their residence time in the plasma.

Optical diagnostics of atmospheric pressure air plasmas

International Nuclear Information System (INIS)

Laux, C O; Spence, T G; Kruger, C H; Zare, R N

2003-01-01

Atmospheric pressure air plasmas are often thought to be in local thermodynamic equilibrium owing to fast interspecies collisional exchange at high pressure. This assumption cannot be relied upon, particularly with respect to optical diagnostics. Velocity gradients in flowing plasmas and/or elevated electron temperatures created by electrical discharges can result in large departures from chemical and thermal equilibrium. This paper reviews diagnostic techniques based on optical emission spectroscopy and cavity ring-down spectroscopy that we have found useful for making temperature and concentration measurements in atmospheric pressure plasmas under conditions ranging from thermal and chemical equilibrium to thermochemical nonequilibrium

The inner containment of an EPR trademark pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

2014-08-15

On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

Thermohydraulic feedbacks in self-pressurized reactor systems

International Nuclear Information System (INIS)

Fiebig, R.

1977-01-01

The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controlable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

Thermohydraulic feedbacks in self-pressurized reactor systems

International Nuclear Information System (INIS)

Fiebig, R.

1977-01-01

The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controllable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

Cascading pressure reactor and method for solar-thermochemical reactions

Science.gov (United States)

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.

Development of a pressurizer level compensator for use on N Reactor

International Nuclear Information System (INIS)

Bussell, J.H.

1985-07-01

The instrument described in this report has been developed to compensate the measured water level in the N Reactor pressurizer for temperature effects. N Reactor is a pressurized water nuclear reactor (PWR). The instrument is defined as a pressurizer level compensator (PLC). A pressurizer is used in a PWR to control the primary coolant pressure and provide a surge volume for primary coolant expansion and contraction. A means of compensating for water and steam density is required because of the wide range of pressure and temperature that result from different steady state and transient reactor power levels. The uncompensated level is determined by measurement of differential pressure between the top of the level measurement zone and the bottom of the level measurement zone. Temperature of the water in the pressurizer is the parameter that is used to determine the proper level compensation since water and steam density are primarily functions of temperature in this case. The PLC uses a microprocessor to calculate the compensated level from temperature and differential pressure measurements. This report includes a description of the design, development, and implementation of software and hardware that are in the PLC. 9 refs., 51 figs., 17 tabs

An investigation of transient pressures and plasma properties in a pinched plasma column. M.S. Thesis

Science.gov (United States)

Stover, E. K.; York, T. M.

1971-01-01

The transient pinched plasma column generated in a linear Z-pinch was studied experimentally and analytically. The plasma column was investigated experimentally with several plasma diagnostics; they were: a rapid response pressure transducer, a magnetic field probe, a voltage probe, and discharge luminosity. Axial pressure profiles on the discharge chamber axis were used to identify three characteristic regions of plasma column behavior: (1) strong axial pressure asymmetry noted early in plasma column lifetime, (2) followed by plasma heating in which there is a rapid rise in static pressure, and (3) a slight decrease static pressure before plasma column breakup. Plasma column lifetime was approximately 5 microseconds. The axial pressure asymmetry was attributed to nonsimultaneous pinching of the imploding current sheet along the discharge chamber axis. The rapid heating could be attributed in part to viscous effects introduced by radial gradients in the axial streaming velocity.

BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

International Nuclear Information System (INIS)

1987-11-01

This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

Converting a Microwave Oven into a Plasma Reactor: A Review

Directory of Open Access Journals (Sweden)

Victor J. Law

2018-01-01

Full Text Available This paper reviews the use of domestic microwave ovens as plasma reactors for applications ranging from surface cleaning to pyrolysis and chemical synthesis. This review traces the developments from initial reports in the 1980s to today’s converted ovens that are used in proof-of-principle manufacture of carbon nanostructures and batch cleaning of ion implant ceramics. Information sources include the US and Korean patent office, peer-reviewed papers, and web references. It is shown that the microwave oven plasma can induce rapid heterogeneous reaction (solid to gas and liquid to gas/solid plus the much slower plasma-induced solid state reaction (metal oxide to metal nitride. A particular focus of this review is the passive and active nature of wire aerial electrodes, igniters, and thermal/chemical plasma catalyst in the generation of atmospheric plasma. In addition to the development of the microwave oven plasma, a further aspect evaluated is the development of methodologies for calibrating the plasma reactors with respect to microwave leakage, calorimetry, surface temperature, DUV-UV content, and plasma ion densities.

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

An Atmospheric Pressure Plasma Setup to Investigate the Reactive Species Formation.

Science.gov (United States)

Gorbanev, Yury; Soriano, Robert; O'Connell, Deborah; Chechik, Victor

2016-11-03

Non-thermal atmospheric pressure ('cold') plasmas have received increased attention in recent years due to their significant biomedical potential. The reactions of cold plasma with the surrounding atmosphere yield a variety of reactive species, which can define its effectiveness. While efficient development of cold plasma therapy requires kinetic models, model benchmarking needs empirical data. Experimental studies of the source of reactive species detected in aqueous solutions exposed to plasma are still scarce. Biomedical plasma is often operated with He or Ar feed gas, and a specific interest lies in investigation of the reactive species generated by plasma with various gas admixtures (O2, N2, air, H2O vapor, etc.) Such investigations are very complex due to difficulties in controlling the ambient atmosphere in contact with the plasma effluent. In this work, we addressed common issues of 'high' voltage kHz frequency driven plasma jet experimental studies. A reactor was developed allowing the exclusion of ambient atmosphere from the plasma-liquid system. The system thus comprised the feed gas with admixtures and the components of the liquid sample. This controlled atmosphere allowed the investigation of the source of the reactive oxygen species induced in aqueous solutions by He-water vapor plasma. The use of isotopically labelled water allowed distinguishing between the species originating in the gas phase and those formed in the liquid. The plasma equipment was contained inside a Faraday cage to eliminate possible influence of any external field. The setup is versatile and can aid in further understanding the cold plasma-liquid interactions chemistry.

An investigation of transient pressure and plasma properties in a pinched plasma column. M.S. Thesis

Science.gov (United States)

Stover, E. K.; York, T. M.

1971-01-01

The transient pinched plasma column generated in a linear Z-pinch was studied experimentally and analytically. The plasma column was investigated experimentally with the following plasma diagnostics: a special rapid response pressure transducer, a magnetic field probe, a voltage probe and discharge luminosity. Axial pressure profiles on the discharge chamber axis were used to identify three characteristic regions of plasma column behavior; they were in temporal sequence: strong axial pressure asymmetry noted early in plasma column lifetime followed by plasma heating in which there is a rapid rise in static pressure and a slight decrease static pressure before plasma column breakup. Plasma column lifetime was approximately 5 microseconds. The axial pressure asymmetry was attributed to nonsimultaneous pinching of the imploding current sheet along the discharge chamber axis. The rapid heating is attributed in part to viscous effects introduced by radial gradients in the axial streaming velocity. Turbulent heating arising from discharge current excitation of the ion acoustic wave instability is also considered a possible heating mechanism.

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Tearing modes with pressure gradient effect in pair plasmas

International Nuclear Information System (INIS)

Cai Huishan; Li Ding; Zheng Jian

2009-01-01

The general dispersion relation of tearing mode with pressure gradient effect in pair plasmas is derived analytically. If the pressure gradients of positron and electron are not identical in pair plasmas, the pressure gradient has significant influence at tearing mode in both collisionless and collisional regimes. In collisionless regime, the effects of pressure gradient depend on its magnitude. For small pressure gradient, the growth rate of tearing mode is enhanced by pressure gradient. For large pressure gradient, the growth rate is reduced by pressure gradient. The tearing mode can even be stabilized if pressure gradient is large enough. In collisional regime, the growth rate of tearing mode is reduced by the pressure gradient. While the positron and electron have equal pressure gradient, tearing mode is not affected by pressure gradient in pair plasmas.

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

Reactor pressure vessel steels

International Nuclear Information System (INIS)

Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

1998-01-01

Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

A dielectric barrier discharge (DBD) plasma reactor: an efficient tool to prepare novel RuO2 nanorods

International Nuclear Information System (INIS)

Ananth, Antony; Gandhi, Mani Sanjeeva; Mok, Young Sun

2013-01-01

One-dimensional (1D) nanostructured materials have attracted a great deal of interest owing to their potential applications in various industries. Due to the limitations and cost associated with conventional low-pressure plasma systems, atmospheric-pressure plasma techniques such as dielectric barrier discharges (DBDs) are investigated as an alternative approach for inducing specific chemical reactions. RuO 2 nanomaterials are widely used as supercapacitor electrodes, in field-emission devices and for catalytic applications. In such applications, size and shape dependent properties of nanomaterials play critical roles in improving the performance. In this paper, an attempt is made to prepare 1D RuO 2 nanostructured materials using a DBD plasma. It is reported here that the composition of feed gas is an important factor in determining the final morphology. For example, an Ar + H 2 plasma yields aggregated RuO 2 nanostructures, whereas ‘nanopillar’ and ‘nanorod’ morphologies are obtained when using Ar + O 2 and Ar, respectively. Possible mechanisms behind the morphological differences are elucidated on the basis of the temperature variations inside the plasma reactor and the chemistry of the gaseous reactive species. The application of a DBD plasma to the synthesis of RuO 2 nanorods is reported for the first time in this paper. (paper)

Renovation of the sealing planes of WWER-400 reactors pressure vessel

International Nuclear Information System (INIS)

Jablonicky, P.; Pilat, P.

2007-01-01

An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

Radiation effects on reactor pressure vessel supports

International Nuclear Information System (INIS)

Johnson, R.E.

1996-05-01

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

Pressure test method for reactor pressure vessel in construction field

International Nuclear Information System (INIS)

Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

1998-01-01

Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

3D Simulation of a Loss of Vacuum Accident (LOVA in ITER (International Thermonuclear Experimental Reactor: Evaluation of Static Pressure, Mach Number, and Friction Velocity

Directory of Open Access Journals (Sweden)

Jean-François Ciparisse

2018-04-01

Full Text Available ITER (International Thermonuclear Experimental Reactor is a magnetically confined plasma nuclear reactor. Inside it, due to plasma disruptions, the formation of neutron-activated powders, which are essentially made out of tungsten and beryllium, occurs. As many windows for diagnostics are present on the reactor, which operates at very low pressure, a LOVA (Loss of Vacuum Accident could be possible and may lead to dust mobilisation and a toxic and radioactive fallout inside the plant. This study is aimed at reproducing numerically the first seconds of a LOVA in ITER, in order to get information about the dust resuspension risk. This work has been carried out by means of a CFD (Computational Fluid Dynamics simulation of the beginning of the pressurisation transient inside the whole Tokamak. It has been found that the pressurization transient is extremely slow, and that the friction speed on the walls is very high, and therefore a high mobilization risk of the dust is expected on the entire internal surface of the reactor. It has been observed that a LOVA in a real-scale reactor is more severe than the one reproduced in reduced-scale facilities, as STARDUST-U, because the speeds are higher, and the dust resuspension capacity of the flow is greater.

The European pressurized water reactor

International Nuclear Information System (INIS)

Leny, J.C.

1993-01-01

The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

Head spray nozzle in reactor pressure vessel

International Nuclear Information System (INIS)

Hatano, Shun-ichi.

1990-01-01

In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

International Nuclear Information System (INIS)

Adamec, P.

2000-12-01

Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

Ultrasound enhanced plasma surface modification at atmospheric pressure

DEFF Research Database (Denmark)

Kusano, Yukihiro; Singh, Shailendra Vikram; Norrman, Kion

and the material surface, and thus many reactive species generated in the plasma can reach the surface before inactivated, and be efficiently utilized for surface modification. In the present work polyester plates are treated using a dielectric barrier discharge (DBD) and a gliding arc at atmospheric pressure......Atmospheric pressure plasma treatment can be highly enhanced by simultaneous high-power ultrasonic irradiation onto the treating surface. It is because ultrasonic waves with a sound pressure level (SPL) above approximately 140 dB can reduce the thickness of a boundary gas layer between the plasma...... irradiation, the water contact angle dropped markedly, and tended to decrease furthermore at higher power. The ultrasonic irradiation during the plasma treatment consistently improved the wettability. Oxygen containing polar functional groups were introduced at the surface by the plasma treatment...

The Use of Multi-Reactor Cascade Plasma Electrolysis for Linear Alkylbenzene Sulfonate Degradation

Science.gov (United States)

Saksono, Nelson; Ibrahim; Zainah; Budikania, Trisutanti

2018-03-01

Plasma electrolysis is a method that can produce large amounts of hydroxyl radicals to degrade organic waste. The purpose of this study is to improve the effectiveness of Linear alkylbenzene sulfonate (LAS) degradation by using multi-reactor cascade plasma electrolysis. The reactor which operated in circulation system, using 3 reactors series flow and 6 L of LAS with initial concentration of 100 ppm. The results show that the LAS degradation can be improved multi-reactor cascade plasma electrolysis. The greatest LAS degradation is achieved up to 81.91% with energy consumption of 2227.34 kJ/mmol that is obtained during 120 minutes by using 600 Volt, 0.03 M of KOH, and 0.5 cm of the anode depth.

The pressurized water reactor

International Nuclear Information System (INIS)

Gallagher, J.L.

1987-01-01

Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

Ultrasound enhanced plasma surface modification at atmospheric pressure

DEFF Research Database (Denmark)

Kusano, Yukihiro; Singh, Shailendra Vikram; Norrman, Kion

2012-01-01

Efficiency of atmospheric pressure plasma treatment can be highly enhanced by simultaneous high power ultrasonic irradiation onto the treating surface. It is because ultrasonic waves with a sound pressure level (SPL) above ∼140 dB can reduce the thickness of a boundary gas layer between the plasma...... arc at atmospheric pressure to study adhesion improvement. The effect of ultrasonic irradiation with the frequency diapason between 20 and 40 kHz at the SPL of ∼150 dB was investigated. After the plasma treatment without ultrasonic irradiation, the wettability was significantly improved...

Pressure thermal shock analysis for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

2015-01-01

The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

Abatement of fluorinated compounds using a 2.45 GHz microwave plasma torch with a reverse vortex plasma reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, J.H.; Cho, C.H.; Shin, D.H. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Hong, Y.C., E-mail: ychong@nfri.re.kr [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Shin, Y.W. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); School of Advanced Green Energy and Environments, Handong Global University, Heunghae-eup, Buk-gu, Pohang-city, Gyeongbuk (Korea, Republic of)

2015-08-30

Highlights: • We developed a microwave plasma torch with reverse vortex reactor (RVR). • We calculated a volume fraction and temperature distribution of discharge gas and waste. • The performance of reverse vortex reactor increased from 29% to 43% than conventional vortex reactor. - Abstract: Abatement of fluorinated compounds (FCs) used in semiconductor and display industries has received an attention due to the increasingly stricter regulation on their emission. We have developed a 2.45 GHz microwave plasma torch with reverse vortex reactor (RVR). In order to design a reverse vortex plasma reactor, we calculated a volume fraction and temperature distribution of discharge gas and waste gas in RVR by ANSYS CFX of computational fluid dynamics (CFD) simulation code. Abatement experiments have been performed with respect to SF{sub 6}, NF{sub 3} by varying plasma power and N{sub 2} flow rates, and FCs concentration. Detailed experiments were conducted on the abatement of NF{sub 3} and SF{sub 6} in terms of destruction and removal efficiency (DRE) using Fourier transform infrared (FTIR). The DRE of 99.9% for NF{sub 3} was achieved without an additive gas at the N{sub 2} flow rate of 150 liter per minute (L/min) by applying a microwave power of 6 kW with RVR. Also, a DRE of SF{sub 6} was 99.99% at the N{sub 2} flow rate of 60 L/min using an applied microwave power of 6 kW. The performance of reverse vortex reactor increased about 43% of NF{sub 3} and 29% of SF{sub 6} abatements results definition by decomposition energy per liter more than conventional vortex reactor.

A conceptual fusion reactor based on the high-plasma-density Z-pinch

International Nuclear Information System (INIS)

Hartman, C.W.; Carlson, G.; Hoffman, M.; Werner, R.

1977-01-01

Conceptual DT and DD fusion reactors are discussed based on magnetic confinement with the high-plasma-density Z-pinch. The reactor concepts have no ''first wall'', the fusion neutrons and plasma energy being absorbed directly into a surrounding lithium vortex blanket. Efficient systems with low re-circulated power are projected, based on a flow-through pinch cycle for which overall Q values can approach 10. The conceptual reactors are characterized by simplicity, small minimum size (100MW(e)) and by the potential for minimal radioactivity hazards. (author)

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Plastic reactor suitable for high pressure and supercritical fluid electrochemistry

DEFF Research Database (Denmark)

Branch, Jack; Alibouri, Mehrdad; Cook, David A.

2017-01-01

The paper describes a reactor suitable for high pressure, particularly supercritical fluid, electrochemistry and electrodeposition at pressures up to 30 MPa at 115◦C. The reactor incorporates two key, new design concepts; a plastic reactor vessel and the use of o-ring sealed brittle electrodes...... by the deposition of Bi. The application of the reactor to the production of nanostructures is demonstrated by the electrodeposition of ∼80 nm diameter Te nanowires into an anodic alumina on silicon template. Key advantages of the new reactor design include reduction of the number of wetted materials, particularly...... glues used for insulating electrodes, compatability with reagents incompatible with steel, compatability with microfabricated planar multiple electrodes, small volume which brings safety advantages and reduced reagent useage, and a significant reduction in experimental time....

Analysis code for pressure in reactor containment vessel of ATR. CONPOL

International Nuclear Information System (INIS)

1997-08-01

For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

Plant life management strategies for pressurized heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

Development and design of control rod drive mechanisms for pressurized water reactors

International Nuclear Information System (INIS)

Leme, Francisco Louzano

2003-01-01

The Control Rod Drive Mechanisms (CRDM) for a Pressurized Water Reactor (PWR) are equipment, integrated to the reactor pressure vessel, incorporating mechanical and electrical components designed to move and position the control rods to guarantee the control of power and shutdown of the nuclear reactor, during normal operation, either in emergency or accidental situations. The type of CRDM used in PWR reactors, whose detailed individual description will be presented in this monograph are the Roller-Nut and Magnetic-Jack. The environment, where the CRDM performs its above presented operational functions, includes direct contact with the fluid used as coolant peculiar to the interior of the reactor, and its associated chemical characteristics, the radiation field next to the reactor core, and also the temperature and pressure in the reactor pressure vessel. So the importance of the CRDM design requirements related to its safety functions are emphasized. Finally, some aspects related to the mechanical and structural design of CRDM of a case study, considering the CRDM for a PWR from the experimental nuclear plant to be applied by CTMSP (Centro Tecnologico da Marinha em Sao Paulo), are pointed out. The design and development of these equipment (author)

The safety of pressurized water reactors

International Nuclear Information System (INIS)

Panossian, J.; Tanguy, P.

1991-01-01

In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Pajnic, M.

2006-01-01

The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

Reactor Pressure Vessel Steels

Energy Technology Data Exchange (ETDEWEB)

Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

1998-07-01

SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

Radiation embrittlement in pressure vessels of power reactors

International Nuclear Information System (INIS)

Kempf, Rodolfo; Fortis, Ana M.

2007-01-01

It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

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

Energy Technology Data Exchange (ETDEWEB)

Beach, N. J.; Knapp, S. M. M.; Landis, C. R., E-mail: landis@chem.wisc.edu [Department of Chemistry, University of Wisconsin-Madison, Madison, Wisconsin 53719 (United States)

2015-10-15

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 to 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.

Pressure vessel codes: Their application to nuclear reactor systems

International Nuclear Information System (INIS)

1966-01-01

A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts

International Nuclear Information System (INIS)

Duffey, R.B.; Pioro, I.L.; Gabaraev, B.A.; Kuznetsov, Yu. N.

2006-01-01

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

Deactivation of Escherichia coli in a post-discharge chamber coupled to an atmospheric pressure multi-electrode DBD plasma source

International Nuclear Information System (INIS)

Pérez-Ruiz, V H; López-Callejas, R; De la Piedad Beneitez, A; Peña-Eguiluz, R; Mercado-Cabrera, A; Muñoz-Castro, A E; Barocio, S R; Valencia-Alvarado, R; Rodríguez-Méndez, B G

2012-01-01

Experimental results from applying a room pressure RF multi-electrode DBD plasma source to the inhibition of the population growth of Gram negative Escherichia coli (E. coli) within a post-discharge reactor are reported. The sample to be treated is deposited in the post-discharge chamber at about 50 mm from the plasma source outlet. Thus, the active species generated by the source are conveyed toward the chamber by the working gas flow. The plasma characterization included the measurement of the axial temperature at different distances from the reactor outlet by means of a K-type thermocouple. The resulting 294 K to 322 K temperature interval corresponded to distances between 10 mm to 1 mm respectively. As the material under treatment is placed further away, any thermal damage of the sample by the plasma is prevented. The measurement and optimization of the ozone O 3 concentration has also been carried out, provided that this is an active specie with particularly high germicide power. The effectiveness treatment of the E. coli bacteria growth inhibition by the proposed plasma source reached 99% when a 10 3 CFU/mL concentration on an agar plate had been exposed during ten minutes.

Welding of the A1 reactor pressure vessel

International Nuclear Information System (INIS)

Becka, J.

1975-01-01

As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

International Nuclear Information System (INIS)

Frid, W.

1988-04-01

Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

A quantitative methodology for reactor vessel pressurized thermal shock decision making

International Nuclear Information System (INIS)

Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

1983-01-01

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

“Virtual IED sensor” at an rf-biased electrode in low-pressure plasma

Energy Technology Data Exchange (ETDEWEB)

Bogdanova, M. A.; Zyryanov, S. M. [Skobeltsyn Institute of Nuclear Physics, Moscow State University, SINP MSU, Moscow (Russian Federation); Faculty of Physics, Moscow State University, MSU, Moscow (Russian Federation); Lopaev, D. V.; Rakhimov, A. T. [Skobeltsyn Institute of Nuclear Physics, Moscow State University, SINP MSU, Moscow (Russian Federation)

2016-07-15

Energy distribution and the flux of the ions coming on a surface are considered as the key-parameters in anisotropic plasma etching. Since direct ion energy distribution (IED) measurements at the treated surface during plasma processing are often hardly possible, there is an opportunity for virtual ones. This work is devoted to the possibility of such indirect IED and ion flux measurements at an rf-biased electrode in low-pressure rf plasma by using a “virtual IED sensor” which represents “in-situ” IED calculations on the absolute scale in accordance with a plasma sheath model containing a set of measurable external parameters. The “virtual IED sensor” should also involve some external calibration procedure. Applicability and accuracy of the “virtual IED sensor” are validated for a dual-frequency reactive ion etching (RIE) inductively coupled plasma (ICP) reactor with a capacitively coupled rf-biased electrode. The validation is carried out for heavy (Ar) and light (H{sub 2}) gases under different discharge conditions (different ICP powers, rf-bias frequencies, and voltages). An EQP mass-spectrometer and an rf-compensated Langmuir probe (LP) are used to characterize plasma, while an rf-compensated retarded field energy analyzer (RFEA) is applied to measure IED and ion flux at the rf-biased electrode. Besides, the pulsed selfbias method is used as an external calibration procedure for ion flux estimating at the rf-biased electrode. It is shown that pulsed selfbias method allows calibrating the IED absolute scale quite accurately. It is also shown that the “virtual IED sensor” based on the simplest collisionless sheath model allows reproducing well enough the experimental IEDs at the pressures when the sheath thickness s is less than the ion mean free path λ{sub i} (s < λ{sub i}). At higher pressure (when s > λ{sub i}), the difference between calculated and experimental IEDs due to ion collisions in the sheath is observed in the low

Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

2009-01-01

An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

Nuclear reactor of pressurized liquid coolant type

International Nuclear Information System (INIS)

Costes, D.

1976-01-01

The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

Particle formation and its control in dual frequency plasma etching reactors

International Nuclear Information System (INIS)

Kim, Munsu; Cheong, Hee-Woon; Whang, Ki-Woong

2015-01-01

The behavior of a particle cloud in plasma etching reactors at the moment when radio frequency (RF) power changes, that is, turning off and transition steps, was observed using the laser-light-scattering method. Two types of reactors, dual-frequency capacitively coupled plasma (CCP) and the hybrid CCP/inductively coupled plasma (ICP), were set up for experiments. In the hybrid CCP/ICP reactor (hereafter ICP reactor), the position and shape of the cloud were strongly dependent on the RF frequency. The particle cloud becomes larger and approaches the electrode as the RF frequency increases. By turning the lower frequency power off later with a small delay time, the particle cloud is made to move away from the electrode. Maintaining lower frequency RF power only was also helpful to reduce the particle cloud size during this transition step. In the ICP reactor, a sufficient bias power is necessary to make a particle trap appear. A similar particle cloud to that in the CCP reactor was observed around the sheath region of the lower electrode. The authors can also use the low-frequency effect to move the particle cloud away from the substrate holder if two or more bias powers are applied to the substrate holder. The dependence of the particle behavior on the RF frequencies suggests that choosing the proper frequency at the right moment during RF power changes can reduce particle contamination effectively

Design of virtual SCADA simulation system for pressurized water reactor

International Nuclear Information System (INIS)

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

2016-01-01

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

Design of virtual SCADA simulation system for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

2016-02-08

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.

Coolant mixing in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Hoehne, T; Grunwald, G

1998-10-01

The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

Problems in manufacturing and transport of pressure vessels of integral reactors

International Nuclear Information System (INIS)

Kralovec, J.

1997-01-01

Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, i.e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters, heights and thick walls and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabilities. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment, machining, shop-handling, non-destructive testing, hydraulic pressure tests etc. Tile transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport each of them having its advantages and limitations. Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper. (author)

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

Hydrophobic and superhydrophobic surfaces fabricated using atmospheric pressure cold plasma technology: A review.

Science.gov (United States)

Dimitrakellis, Panagiotis; Gogolides, Evangelos

2018-04-01

Hydrophobic surfaces are often used to reduce wetting of surfaces by water. In particular, superhydrophobic surfaces are highly desired for several applications due to their exceptional properties such as self-cleaning, anti-icing, anti-friction and others. Such surfaces can be prepared via numerous methods including plasma technology, a dry technique with low environmental impact. Atmospheric pressure plasma (APP) has recently attracted significant attention as lower-cost alternative to low-pressure plasmas, and as a candidate for continuous rather than batch processing. Although there are many reviews on water-repellent surfaces, and a few reviews on APP technology, there are hardly any review works on APP processing for hydrophobic and superhydrohobic surface fabrication, a topic of high importance in nanotechnology and interface science. Herein, we critically review the advances on hydrophobic and superhydrophobic surface fabrication using APP technology, trying also to give some perspectives in the field. After a short introduction to superhydrophobicity of nanostructured surfaces and to APPs we focus this review on three different aspects: (1) The atmospheric plasma reactor technology used for fabrication of (super)hydrophobic surfaces. (2) The APP process for hydrophobic surface preparation. The hydrophobic surface preparation processes are categorized methodologically as: a) activation, b) grafting, c) polymerization, d) roughening and hydrophobization. Each category includes subcategories related to different precursors used. (3) One of the most important sections of this review concerns superhydrophobic surfaces fabricated using APP. These are methodologically characterized as follows: a) single step processes where micro-nano textured topography and low surface energy coating are created at the same time, or b) multiple step processes, where these steps occur sequentially in or out of the plasma. We end the review with some perspectives in the field. We

Atmospheric-pressure plasma activation and surface characterization on polyethylene membrane separator

Science.gov (United States)

Tseng, Yu-Chien; Li, Hsiao-Ling; Huang, Chun

2017-01-01

The surface hydrophilic activation of a polyethylene membrane separator was achieved using an atmospheric-pressure plasma jet. The surface of the atmospheric-pressure-plasma-treated membrane separator was found to be highly hydrophilic realized by adjusting the plasma power input. The variations in membrane separator chemical structure were confirmed by Fourier transform infrared spectroscopy and X-ray photoelectron spectroscopy. Chemical analysis showed newly formed carbonyl-containing groups and high surface concentrations of oxygen-containing species on the atmospheric-pressure-plasma-treated polymeric separator surface. It also showed that surface hydrophilicity primarily increased from the polar component after atmospheric-pressure plasma treatment. The surface and pore structures of the polyethylene membrane separator were examined by scanning electron microscopy, revealing a slight alteration in the pore structure. As a result of the incorporation of polar functionalities by atmospheric-pressure plasma activation, the electrolyte uptake and electrochemical impedance of the atmospheric-pressure-plasma-treated membrane separator improved. The investigational results show that the separator surface can be controlled by atmospheric-pressure plasma surface treatment to tailor the hydrophilicity and enhance the electrochemical performance of lithium ion batteries.

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

International Nuclear Information System (INIS)

Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

Rupture tests with reactor pressure vessel head models

International Nuclear Information System (INIS)

Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

2003-01-01

In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

Novel Composite Hydrogen-Permeable Membranes for Non-Thermal Plasma Reactors for the Decomposition of Hydrogen Sulfide

Energy Technology Data Exchange (ETDEWEB)

Morris D. Argyle; John F. Ackerman; Suresh Muknahallipatna; Jerry C. Hamann; Stanislaw Legowski; Guibling Zhao; Ji-Jun Zhang; Sanil John

2005-10-01

The goal of this experimental project is to design and fabricate a reactor and membrane test cell to dissociate hydrogen sulfide (H{sub 2}S) in a non-thermal plasma and recover hydrogen (H{sub 2}) through a superpermeable multi-layer membrane. Superpermeability of hydrogen atoms (H) has been reported by some researchers using membranes made of Group V transition metals (niobium, tantalum, vanadium, and their alloys), although it has yet to be confirmed in this study. A pulsed corona discharge (PCD) reactor has been fabricated and used to dissociate H{sub 2}S into hydrogen and sulfur. A nonthermal plasma cannot be produced in pure H{sub 2}S with our reactor geometry, even at discharge voltages of up to 30 kV, because of the high dielectric strength of pure H{sub 2}S ({approx}2.9 times higher than air). Therefore, H{sub 2}S was diluted in another gas with lower breakdown voltage (or dielectric strength). Breakdown voltages of H{sub 2}S in four balance gases (Ar, He, N{sub 2} and H{sub 2}) have been measured at different H{sub 2}S concentrations and pressures. Breakdown voltages are proportional to the partial pressure of H{sub 2}S and the balance gas. H{sub 2}S conversion and the reaction energy efficiency depend on the balance gas and H{sub 2}S inlet concentrations. With increasing H{sub 2}S concentrations, H{sub 2}S conversion initially increases, reaches a maximum, and then decreases. H{sub 2}S conversion in atomic balance gases, such as Ar and He, is more efficient than that in diatomic balance gases, such as N{sub 2} and H{sub 2}. These observations can be explained by the proposed reaction mechanism of H{sub 2}S dissociation in different balance gases. The results show that nonthermal plasmas are effective for dissociating H{sub 2}S into hydrogen and sulfur.

Development of alternative fuel for pressurized water reactors

International Nuclear Information System (INIS)

Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

1984-01-01

The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

Modelling of an intermediate pressure microwave oxygen discharge reactor: from stationary two-dimensional to time-dependent global (volume-averaged) plasma models

International Nuclear Information System (INIS)

Kemaneci, Efe; Graef, Wouter; Rahimi, Sara; Van Dijk, Jan; Kroesen, Gerrit; Carbone, Emile; Jimenez-Diaz, Manuel

2015-01-01

A microwave-induced oxygen plasma is simulated using both stationary and time-resolved modelling strategies. The stationary model is spatially resolved and it is self-consistently coupled to the microwaves (Jimenez-Diaz et al 2012 J. Phys. D: Appl. Phys. 45 335204), whereas the time-resolved description is based on a global (volume-averaged) model (Kemaneci et al 2014 Plasma Sources Sci. Technol. 23 045002). We observe agreement of the global model data with several published measurements of microwave-induced oxygen plasmas in both continuous and modulated power inputs. Properties of the microwave plasma reactor are investigated and corresponding simulation data based on two distinct models shows agreement on the common parameters. The role of the square wave modulated power input is also investigated within the time-resolved description. (paper)

Magnetic pressure effects in a plasma-liner interface

Science.gov (United States)

García-Rubio, F.; Sanz, J.

2018-04-01

A theoretical analysis of magnetic pressure effects in a magnetized liner inertial fusion-like plasma is presented. In previous publications [F. García-Rubio and J. Sanz, Phys. Plasmas 24, 072710 (2017)], the evolution of a hot magnetized plasma in contact with a cold unmagnetized plasma, aiming to represent the hot spot and liner, respectively, was investigated in planar geometry. The analysis was made in a double limit low Mach and high thermal to magnetic pressure ratio β. In this paper, the analysis is extended to an arbitrary pressure ratio. Nernst, Ettingshausen, and Joule effects come into play in the energy balance. The region close to the liner is governed by thermal conduction, while the Joule dissipation becomes predominant far from it when the pressure ratio is low. Mass ablation, thermal energy, and magnetic flux losses are reduced with plasma magnetization, characterized by the electron Hall parameter ω e τ e , until β values of order unity are reached. From this point forward, increasing the electron Hall parameter no longer improves the magnetic flux conservation, and mass ablation is enhanced due to the magnetic pressure gradients. A thoughtful simplification of the problem that allows to reduce the order of the system of governing equations while still retaining the finite β effects is presented and compared to the exact case.

Origin of fluctuations in atmospheric pressure arc plasma devices

International Nuclear Information System (INIS)

Ghorui, S.; Das, A.K.

2004-01-01

Fluctuations in arc plasma devices are extremely important for any technological application in thermal plasma. The origin of such fluctuations remains unexplained. This paper presents a theory for observed fluctuations in atmospheric pressure arc plasma devices. A qualitative explanation for observed behavior on atmospheric pressure arc plasma fluctuations, reported in the literature, can be obtained from the theory. The potential of the theory is demonstrated through comparison of theoretical predictions with reported experimental observations

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

International Nuclear Information System (INIS)

Shen Yongjun; Ding Jiandong; Lei Lecheng; Zhang Xingwang

2014-01-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10 −9 mol/L and 0.61 × 10 −9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10 −2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10 −2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation

Dismantling of JPDR reactor internals by underwater plasma arc cutting technique using robotic manipulator

International Nuclear Information System (INIS)

Yokota, M.

1988-01-01

The actual dismantling of JPDR started on December 4, 1986. As of now, equipment that surrounds the reactor has mostly been removed to provide working space in reactor containment prior to the dismantling of reactor internals. Some reactor internals have been successfully dismantled using the underwater arc cutting system with a robotic manipulator during the period of January to March 1988. The cutting system is composed of an underwater plasma arc cutting device and a robotic manipulator. The cut off reactor internals were core spray block, feedwater sparger and stabilizers for fuel upper grid tube. The plasma arc cutting device was developed to dismantle the reactor internals underwater. It mainly consists of a plasma torch, power and gas supply systems for the torch, and by-product treatment systems. It has the cutting ability of 130 mm thickness stainless steel underwater. The robotic manipulator has seven degrees of freedom of movement, enabling it to move in almost the same way as the arm of a human being. The arm of the robot is mounted on a supporting device which is suspended by three chains from the support structure set on a service floor. A plasma torch is griped by the robotic hand; its position to the structure to be cut is controlled from a remote control room, about 100 meters outside the reactor containment

Leak detection device for nuclear reactor pressure vessel

International Nuclear Information System (INIS)

Ikeda, Jun.

1988-01-01

Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

Research on atmospheric pressure plasma processing sewage

Science.gov (United States)

Song, Gui-cai; Na, Yan-xiang; Dong, Xiao-long; Sun, Xiao-liang

2013-08-01

The water pollution has become more and more serious with the industrial progress and social development, so it become a worldwide leading environmental management problem to human survival and personal health, therefore, countries are looking for the best solution. Generally speaking, in this paper the work has the following main achievements and innovation: (1) Developed a new plasma device--Plasma Water Bed. (2) At atmospheric pressure condition, use oxygen, nitrogen, argon and helium as work gas respectively, use fiber spectrometer to atmospheric pressure plasma discharge the emission spectrum of measurement, due to the different work gas producing active particle is different, so can understand discharge, different particle activity, in the treatment of wastewater, has the different degradation effects. (3) Methyl violet solution treatment by plasma water bed. Using plasma drafting make active particles and waste leachate role, observe the decolorization, measurement of ammonia nitrogen removal.

Superheated steam annealing of pressurized water reactor vessel

International Nuclear Information System (INIS)

Porowski, J.S.

1993-01-01

Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

Acoustic Emission for on-line reactor pressure boundary monitoring

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

1985-01-01

The program objective is to develop AE for continuous surveillance to assess flaw growth in reactor pressure boundaries. Technology in the laboratory is being evaluated on structures. Results have demonstrated basic feasibility of the program objective. AE monitoring a long term fatigue test of a pressure vessel demonstrated an instrument system, and the ability to detect unexpected as well as well as known fatigue cracks. Monitoring a nuclear reactor system shows that the coolant flow noise problem is manageable and AE can be detected under simulated operating conditions

Basic conceptions for reactor pressure vessel manipulators and their evaluation

International Nuclear Information System (INIS)

Popp, P.

1987-01-01

The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

Absorber rod bundle actuator in a pressurized water nuclear reactor

International Nuclear Information System (INIS)

Martin, J.; Peletan, R.

1984-01-01

The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

Study of the boron homogenizing process employing an experimental low-pressure bench simulating the IRIS reactor pressurizer – Part I

International Nuclear Information System (INIS)

Bezerra, Jair de Lima; Lira, Carlos Alberto Brayner de Oliveira; Barroso, Antonio Carlos de Oliveira; Lima, Fernando Roberto de Andrade; Bezerra da Silva, Mário Augusto

2013-01-01

Highlights: ► Experimental bench with test section made of transparent acrylic, simulating the pressurizer reactor IRIS. ► Workbench used to study the process of homogenization of boron in the pressurizer IRIS nuclear reactor. ► Results were obtained through videos and digital photos of the test section. - Abstract: The reactivity control of a nuclear reactor to pressurized water is made by means of controlling bars or by boron dilution in the water from the coolant of a primary circuit. The control with boron dilution has great importance, despite inserting small variations in the reactivity in the reactor, as it does not significantly affect the distribution of the neutron flux. A simplified experimental bench with a test section manufactured in transparent acrylic, was built in reduced scale as to be used in a boron homogenizing process, simulating an IRIS reactor pressurizer (International Reactor Innovative and Secure). The bench was assembled in the Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE), an entity linked to the Comissão Nacional de Energia Nuclear (CNEN), Recife – PE

Study of the boron homogenizing process employing an experimental low-pressure bench simulating the IRIS reactor pressurizer – Part II

International Nuclear Information System (INIS)

Bezerra, Jair de Lima; Lira, Carlos Alberto Brayner de Oliveira; Barroso, Antonio Carlos de Oliveira; Lima, Fernando Roberto de Andrade; Silva, Mário Augusto Bezerra da

2013-01-01

Highlights: • Experimental bench with test section made of transparent acrylic, simulating the pressurizer reactor IRIS. • Workbench used to study the process of homogenization of boron in the pressurizer IRIS nuclear reactor. • Results were obtained through videos and digital photos of the test section. - Abstract: The reactivity control of a nuclear reactor to pressurized water is made by means of controlling bars or by boron dilution in the water from the coolant of a primary circuit. The control with boron dilution has great importance, despite inserting small variations in the reactivity in the reactor, as it does not significantly affect the distribution of the neutron flux. A simplified experimental bench with a test section manufactured in transparent acrylic, was built in reduced scale as to be used in a boron homogenizing process, simulating an IRIS reactor pressurizer (International Reactor Innovative and Secure). The bench was assembled in the Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE), an entity linked to the Comissão Nacional de Energia Nuclear (CNEN), Recife–PE

Atmospheric Pressure Plasma Processing for Polymer Adhesion: A Review

DEFF Research Database (Denmark)

Kusano, Yukihiro

2014-01-01

Atmospheric pressure plasma processing has attracted significant interests over decades due to its usefulness and a variety of applications. Adhesion improvement of polymer surfaces is among the most important applications of atmospheric pressure plasma treatment. Reflecting recent significant de...

Self-pressurization analysis of the natural circulation integral nuclear reactor using a new dynamic model

Directory of Open Access Journals (Sweden)

Ali Farsoon Pilehvar

2018-06-01

Full Text Available Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established. Keywords: Condensation Power, Flashing Phenomenon, Natural Circulation, Self-Pressurization, Small Modular Reactor

Modelling of SOEC-FT reactor: Pressure effects on methanation process

International Nuclear Information System (INIS)

Chen, Bin; Xu, Haoran; Ni, Meng

2017-01-01

Highlights: • Numerical study on combined SOEC-FT reactor in pressurized condition. • Effects of operating pressure on co-electrolysis and CH_4 production are studied. • The lower limit temperature of the FT section is dependent on the operating pressure. • The CH_4 production can be improved at higher voltage due to the current increase. • Effects of higher exchange current density is predicted at different temperature. - Abstract: In this paper a numerical model is developed for a novel reactor combining a Solid Oxide Electrolyzer Cell (SOEC) section with a Fischer Tropsch like section for methane production under pressurized & temperature-gradient condition. Governing equations for mass, momentum, charge transport are solved with Finite Element Method. The chemical reaction kinetics of reversible water gas shift reaction and reversible methanation reaction in Ni/YSZ cathode are fully considered. The model is validated by comparing simulation results with experimental data. Parametric simulations are conducted to understand the physical-chemical processes in the reactor with a focus on the pressure effect. It is predicted that the optimal operating pressure is around 3 bar, beyond which the CH_4 conversion ratio (2.5 times enhanced than 1 bar operating) cannot be further increased. It is also found that it is feasible to operate the pressurized SOEC at a lower temperature for CH_4 production with improved catalyst activity.

Use of superheated steam to anneal the reactor pressure vessel

International Nuclear Information System (INIS)

Porowski, J.S.

1994-01-01

Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 degrees F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 degrees F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors

Study of a low power plasma reactor for the synthesis of zinc oxide as window layers in Cu(In,Ga)Se{sub 2} solar cells

Energy Technology Data Exchange (ETDEWEB)

Ma, Alexandre, E-mail: alexadre.ma@chimie-paristech.fr [Institut de Recherche de Chimie Paris (IRCP), Equipe 2PM (Procédés, Plasmas, Microsystèmes), UMR 8247, Chimie ParisTech-CNRS, 11 Rue Pierre et Marie Curie, 75005 Paris (France); Institute of Research and Development of Photovoltaic Energy (IRDEP), UMR7174, EDF-CNRS-Chimie ParisTech, 6 Quai Watier, 78401 Chatou (France); ADEME (French Environment and Energy Management Agency), 20 avenue du Grésillé, BP 90406, 49004 Angers Cedex 01 (France); Donsanti, Frédérique [Institute of Research and Development of Photovoltaic Energy (IRDEP), UMR7174, EDF-CNRS-Chimie ParisTech, 6 Quai Watier, 78401 Chatou (France); Rousseau, Frédéric; Morvan, Daniel [Institut de Recherche de Chimie Paris (IRCP), Equipe 2PM (Procédés, Plasmas, Microsystèmes), UMR 8247, Chimie ParisTech-CNRS, 11 Rue Pierre et Marie Curie, 75005 Paris (France)

2015-05-01

The low power plasma reactor is an original process which allows the deposition of ZnO controlled thickness films from an aqueous precursor solution in a cold plasma only in order to reach high growth rates. The quality of the deposited material (purity, crystallinity, size of the grains…) depends primarily on the interactions in the reactor between the droplets and the plasma. In this study, the parameters of the deposition (composition of the gases, pressure, power, temperature…) were studied and controlled. The doping characteristics are mainly influenced by the concentration of the precursors in the solution and by the method of injection. The final optimizations allowed high growth rates ranging from 0.6 to 1 nm/s. X-ray Diffraction results show a good crystallinity of the deposited layer (würtzite structure). According to the transmittance measurements, the films present a good transparency and a calculated optical gap value ranging between 3.2 and 3.3 eV. This deposition technique using plasma is fast, flexible, low power consuming and easily adaptable. Cu(In,Ga)Se{sub 2} solar cells with a ZnO window layer have successfully been achieved by using the low power plasma reactor and their performances show an efficiency of 11%. - Highlights: • The low power plasma reactor is a very fast and flexible deposition process. • ZnO films are realized from an aqueous nitrate solution in a plasma discharge only. • ZnO films exhibit good properties for the window layer application. • CIGS solar cells present promising performances with an efficiency of 11%.

Manipulator for testing a top-opened reactor pressure vessel

International Nuclear Information System (INIS)

Bauer, R.; Kastl, H.

1991-01-01

The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

Development of the reactor lithium ampoule device for research of spectral-luminescent characteristics of nuclear-excited plasma

Energy Technology Data Exchange (ETDEWEB)

Batyrbekov, E.G. [National Nuclear Center of RK, Kurchatov (Kazakhstan); Gordienko, Yu. N., E-mail: gordienko@nnc.kz [National Nuclear Center of RK, Kurchatov (Kazakhstan); Ponkratov, Yu. V. [National Nuclear Center of RK, Kurchatov (Kazakhstan); Khasenov, M.U. [PI “National Laboratory Astana”, Astana (Kazakhstan); Tazhibayeva, I.L.; Barsukov, N.I.; Kulsartov, T.V.; Zaurbekova, Zh. A.; Tulubayev, Ye. Yu.; Skakov, M.K. [National Nuclear Center of RK, Kurchatov (Kazakhstan)

2017-04-15

Highlights: • The development procedure of the ampoule device for experiments with nuclear-excited plasma under neutron irradiation is described. • The methods of nuclear reactions’ energy conversion into the energy of optical radiation of nuclear-excited plasma are presented. • A scheme of reactor experiments, the experimental facility and experimental device to carry out the reactor experiments are considered. - Abstract: This paper describes the development procedure of the reactor ampoule device to perform the experiments on study of spectral luminescence characteristics of nuclear-excited plasma formed by products of {sup 6}Li(n,α){sup 3}H reaction under neutron irradiation at the IVG.1 M research reactor. The methods of nuclear reactions’ energy conversion into the energy of optical radiation of nuclear-excited plasma are presented. A scheme of reactor experiments, the experimental facility and experimental device to carry out the reactor experiments are considered in paper. The designed ampoule device is totally meets the requirements of irradiation experiments on the IVG.1M reactor.

Predictions of ion energy distributions and radical fluxes in radio frequency biased inductively coupled plasma etching reactors

Science.gov (United States)

Hoekstra, Robert J.; Kushner, Mark J.

1996-03-01

Inductively coupled plasma (ICP) reactors are being developed for low gas pressure (radio frequency (rf) bias is applied to the substrate. One of the goals of these systems is to independently control the magnitude of the ion flux by the inductively coupled power deposition, and the acceleration of ions into the substrate by the rf bias. In high plasma density reactors the width of the sheath above the wafer may be sufficiently thin that ions are able to traverse it in approximately 1 rf cycle, even at 13.56 MHz. As a consequence, the ion energy distribution (IED) may have a shape typically associated with lower frequency operation in conventional reactive ion etching tools. In this paper, we present results from a computer model for the IED incident on the wafer in ICP etching reactors. We find that in the parameter space of interest, the shape of the IED depends both on the amplitude of the rf bias and on the ICP power. The former quantity determines the average energy of the IED. The latter quantity controls the width of the sheath, the transit time of ions across the sheath and hence the width of the IED. In general, high ICP powers (thinner sheaths) produce wider IEDs.

On OH production in water containing atmospheric pressure plasmas

NARCIS (Netherlands)

Bruggeman, P.J.; Schram, D.C.

2010-01-01

In this paper radical production in atmospheric pressure water containing plasmas is discussed. As OH is often an important radical in these discharges the paper focuses on OH production. Besides nanosecond pulsed coronas and diffusive glow discharges, several other atmospheric pressure plasmas

Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

International Nuclear Information System (INIS)

Bozoki, G.; Kohut, P.; Fitzpatrick, R.

1989-02-01

This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

TORT application in reactor pressure vessel neutron flux calculations

International Nuclear Information System (INIS)

Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

1994-01-01

The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2012-02-15

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor

Science.gov (United States)

Abedi-Varaki, Mehdi

2017-08-01

Carbon dioxide (CO2) as the primary greenhouse gas, is the main pollutant that is warming earth. CO2 is widely emitted through the cars, planes, power plants and other human activities that involve the burning of fossil fuels (coal, natural gas and oil). Thus, there is a need to develop some method to reduce CO2 emission. To this end, this study investigates the behavior of CO2 in dielectric barrier discharge (DBD) plasma reactor. The behavior of different species and their reaction rates are studied using a zero-dimensional model based on equations of kinetics inside plasma reactor. The results show that the plasma reactor has an effective reduction on the CO2 density inside the reactor. As a result of reduction in the temporal variations of reaction rate, the speed of chemical reactions for CO2 decreases and very low concentration of CO2 molecules inside the plasma reactor is generated. The obtained results are compared with the existing experimental and simulation findings in the literature.

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

Fueling moving ring field-reversed mirror reactor plasmas

International Nuclear Information System (INIS)

Felber, F.S.

1980-01-01

The concept of small fusion reactors is being studied jointly by Lawrence Livermore Laboratory General Atomic Company, and Pacific Gas and Electric Company. The objective is to investigate alternatives and then to develop a conceptual design for a small reactor that could produce useful, though not necessarily economical, energy by the late 1980s. Three methods of fueling a small moving ring field-reversed mirror are considered: injection of fuel pellets accelerated by laser ablation, injection of fuel pellets accelerated by deflagration-gun ablation, and direct injection of plasma by a deflagration gun. 13 refs

ORNL TNS Program: plasma engineering considerations and innovations for a medium field tokamak fusion reactor

International Nuclear Information System (INIS)

Peng, Y.K.M.; Attenberger, S.E.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

1977-12-01

Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments

Quantum cascade laser absorption spectroscopy with the amplitude-to-time conversion technique for atmospheric-pressure plasmas

International Nuclear Information System (INIS)

Yumii, Takayoshi; Kimura, Noriaki; Hamaguchi, Satoshi

2013-01-01

The NO 2 concentration, i.e., density, in a small plasma of a nitrogen oxide (NOx) treatment reactor has been measured by highly sensitive laser absorption spectroscopy. The absorption spectroscopy uses a single path of a quantum cascade laser beam passing through a plasma whose dimension is about 1 cm. The high sensitivity of spectroscopy is achieved by the amplitude-to-time conversion technique. Although the plasma reactor is designed to convert NO in the input gas to NO 2 , it has been demonstrated by this highly sensitive absorption spectroscopy that NO 2 in a simulated exhaust gas that enters the reactor is decomposed by the plasma first and then NO 2 is formed again, possibly more than it was decomposed, through a series of gas-phase reactions by the time the gas exits the reactor. The observation is consistent with that of an earlier study on NO decomposition by the same type of a plasma reactor [T. Yumii et al., J. Phys. D 46, 135202 (2013)], in which a high concentration of NO 2 was observed at the exit of the reactor.

First results on nitriding aluminium alloys in a low-pressure RF plasma

International Nuclear Information System (INIS)

Fewell, M.P.; Priest, J.M.; Collins, G.A.; Short, K.T.

2000-01-01

Full text: Aluminium alloys are now well established as materials of choice for many commercial applications, especially where strength-to-weight ratio is a critical parameter. However, their more widespread use is inhibited by their low surface hardness. For steels, similar problems can be overcome by nitriding. The nitrogen-rich surface layer has high hardness and load-bearing capacity, and is very well bonded to the substrate. The development of a similar surface-treatment process for aluminium alloys is clearly a desirable goal. It is therefore not surprising that many research groups worldwide have attempted to nitride aluminium. Much of this work studied pure aluminium, a material of no interest for structural applications. Previous investigations into nitriding aluminium alloys' had indifferent results. However, they have served to identify the key issues, which are the importance of a pre-cleaning steps to remove the surface oxide, of impurity control during the nitriding and the desirability of using as low a process temperature as possible. In all of these areas, our process using a low-pressure RF plasma is likely to be competitive. In view of this, we have undertaken a comparative study of a range of commercially available aluminium alloys. All treatments were carried out in the hot-wall nitriding reactor at ANSTO. The samples consist of disks 25mm in diameter and ∼3mm thick which were polished and ultrasonically cleaned in alcohol prior to treatment. The samples were stored in air at all times except when in the nitriding reactor. In a series of treatments, the treatment time was varied in the range 1-16 h and the temperature in the range 350-500 deg C. All treatments were preceeded by a plasma cleaning step in a H 2 /50%Ar mixture for a duration of 1.5-2.0 h while the reactor reached processing temperature. The treatments all used pure N 2 at a pressure of 0.4Pa and a nitrogen flow rate of 12μmol s -1 , with 245W of rf power at 13.56MHz applied to

Some local dilution transient in a pressurized water reactor

International Nuclear Information System (INIS)

Jacobson, S.

1989-01-01

Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

Cavity temperature and flow characteristics in a gas-core test reactor

Science.gov (United States)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

International Nuclear Information System (INIS)

Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon

2012-01-01

Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

Electrochemical noise measurements under pressurized water reactor conditions

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2000-01-01

Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

Non-equilbrium behavior of low-pressure plasma jets

International Nuclear Information System (INIS)

Chang, C.H.; Pfender, E.

1989-01-01

After establishing the basic equations, some sample calculations are presented to examine the thermodynamic state of the plasma from atmospheric to low pressures (80 mbar). These results indicate the validity of local thermodynamic equilibrium (LTE) at atmospheric pressure as well as strong deviations from LTE at lower pressures especially in terms of chemical equilibrium. Departures from kinetic equilibrium are not as severe as those from chemical equilibrium along the centerline of the jet. However, there are some departures from transitional equilibrium in the fringes of the jet. It is demonstrated that conventional methods based on the LTE assumption are not appropriate for describing low-pressure plasma jets

Development, diagnostic and applications of radio-frequency plasma reactor

Science.gov (United States)

Puac, N.

2008-07-01

In many areas of the industry, plasma processing of materials is a vital technology. Nonequilibrium plasmas proved to be able to produce chemically reactive species at a low gas temperature while maintaining highly uniform reaction rates over relatively large areas (Makabe and Petrovic 2006). At the same time nonequilibrium plasmas provide means for good and precise control of the properties of active particles that determine the surface modification. Plasma needle is one of the atmospheric pressure sources that can be used for treatment of the living matter which is highly sensitive when it comes to low pressure or high temperatures (above 40 C). Dependent on plasma conditions, several refined cell responses are induced in mammalian cells (Sladek et al. 2005). It appears that plasma treatment may find many biomedical applications. However, there are few data in the literature about plasma effects on plant cells and tissues. So far, only the effect of low pressure plasmas on seeds was investigated. It was shown that short duration pretreatments by non equilibrium low temperature air plasma were stimulative in light induced germination of Paulownia tomentosa seeds (Puac et al. 2005). As membranes of plants have different properties to those of animals and as they show a wide range of properties we have tried to survey some of the effects of typical plasma which is envisaged to be used in biotechnological applications on plant cells. In this paper we will make a comparison between two configurations of plasma needle that we have used in treatment of biological samples (Puac et al. 2006). Difference between these two configurations is in the additional copper ring that we have placed around glass tube at the tip of the needle. We will show some of the electrical characteristics of the plasma needle (with and without additional copper ring) and, also, plasma emission intensity obtained by using fast ICCD camera.

To the probe theory in a highly-ionized high-pressure plasma

International Nuclear Information System (INIS)

Baksht, F.G.; Rybakov, A.B.

1997-01-01

The probe theory in highly-ionized high-pressure plasma is presented. The situation typical for high-pressure plasma, when the plasma in the main part of the near-probe layer is in the state of local ionization equilibrium with general temperature for electrons and heavy particles. Possibility is discussed for determining the parameters of non-perturbed plasma through analysis of the probe characteristic at place of ion saturation, transition area and by the probe floating potential. The experiments were carried out by example of highly-ionized xenon plasma under atmospheric pressure

Computer simulations of an oxygen inductively coupled plasma used for plasma-assisted atomic layer deposition

International Nuclear Information System (INIS)

Tinck, S; Bogaerts, A

2011-01-01

In this paper, an O 2 inductively coupled plasma used for plasma enhanced atomic layer deposition of Al 2 O 3 thin films is investigated by means of modeling. This work intends to provide more information about basic plasma properties such as species densities and species fluxes to the substrate as a function of power and pressure, which might be hard to measure experimentally. For this purpose, a hybrid model developed by Kushner et al is applied to calculate the plasma characteristics in the reactor volume for different chamber pressures ranging from 1 to 10 mTorr and different coil powers ranging from 50 to 500 W. Density profiles of the various oxygen containing plasma species are reported as well as fluxes to the substrate under various operating conditions. Furthermore, different orientations of the substrate, which can be placed vertically or horizontally in the reactor, are taken into account. In addition, special attention is paid to the recombination process of atomic oxygen on the different reactor walls under the stated operating conditions. From this work it can be concluded that the plasma properties change significantly in different locations of the reactor. The plasma density near the cylindrical coil is high, while it is almost negligible in the neighborhood of the substrate. Ion and excited species fluxes to the substrate are found to be very low and negligible. Finally, the orientation of the substrate has a minor effect on the flux of O 2 , while it has a significant effect on the flux of O. In the horizontal configuration, the flux of atomic oxygen can be up to one order of magnitude lower than in the vertical configuration.

Charge dependence of the plasma travel length in atmospheric-pressure plasma

Energy Technology Data Exchange (ETDEWEB)

Yambe, Kiyoyuki; Konda, Kohmei; Masuda, Seiya [Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan)

2016-06-15

Plasma plume is generated using a quartz tube, helium gas, and foil electrode by applying AC high voltage under the atmosphere. The plasma plume is released into the atmosphere from inside of the quartz tube and is seen as the continuous movement of the plasma bullet. The travel length of plasma bullet is defined from plasma energy and force due to electric field. The drift velocity of plasma bullet has the upper limit under atmospheric-pressure because the drift velocity is determined from the balance between electric field and resistive force due to collisions between plasma and air. The plasma plume charge depends on the drift velocity. Consequently, in the laminar flow of helium gas flow state, the travel length of the plasma plume logarithmically depends on the plasma plume charge which changes with both the electric field and the resistive force.

Charge dependence of the plasma travel length in atmospheric-pressure plasma

International Nuclear Information System (INIS)

Yambe, Kiyoyuki; Konda, Kohmei; Masuda, Seiya

2016-01-01

Plasma plume is generated using a quartz tube, helium gas, and foil electrode by applying AC high voltage under the atmosphere. The plasma plume is released into the atmosphere from inside of the quartz tube and is seen as the continuous movement of the plasma bullet. The travel length of plasma bullet is defined from plasma energy and force due to electric field. The drift velocity of plasma bullet has the upper limit under atmospheric-pressure because the drift velocity is determined from the balance between electric field and resistive force due to collisions between plasma and air. The plasma plume charge depends on the drift velocity. Consequently, in the laminar flow of helium gas flow state, the travel length of the plasma plume logarithmically depends on the plasma plume charge which changes with both the electric field and the resistive force.

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

International Nuclear Information System (INIS)

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-01

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically in influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the part of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material inter actions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (author)

Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

International Nuclear Information System (INIS)

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-01

The major increase in discharge duration and plasma energy in a next-step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several cm from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (orig.)

Pressure balance inconsistency exhibited in a statistical model of magnetospheric plasma

Science.gov (United States)

Garner, T. W.; Wolf, R. A.; Spiro, R. W.; Thomsen, M. F.; Korth, H.

2003-08-01

While quantitative theories of plasma flow from the magnetotail to the inner magnetosphere typically assume adiabatic convection, it has long been understood that these convection models tend to overestimate the plasma pressure in the inner magnetosphere. This phenomenon is called the pressure crisis or the pressure balance inconsistency. In order to analyze it in a new and more detailed manner we utilize an empirical model of the proton and electron distribution functions in the near-Earth plasma sheet (-50 RE attributed to gradient/curvature drift for large isotropic energy invariants but not for small invariants. The tailward gradient of the distribution function indicates a violation of the adiabatic drift condition in the plasma sheet. It also confirms the existence of a "number crisis" in addition to the pressure crisis. In addition, plasma sheet pressure gradients, when crossed with the gradient of flux tube volume computed from the [1989] magnetic field model, indicate Region 1 currents on the dawn and dusk sides of the outer plasma sheet.

Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

International Nuclear Information System (INIS)

Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

1983-07-01

The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

International Nuclear Information System (INIS)

1998-10-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

International Nuclear Information System (INIS)

Hancox, R.; Mitchell, J.T.D.

1977-01-01

Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

Reactor costs and maintenance, with reference to the Culham Mark II conceptual Tokamak reactor design

International Nuclear Information System (INIS)

Hancox, R.; Mitchell, J.T.D.

1976-01-01

Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, due mainly to the low power density of the fusion reactor which affects both the reactor and building costs. In order to reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (orig.) [de

Convective equilibrium and mixing-length theory for stellarator reactors

International Nuclear Information System (INIS)

Ho, D.D.M.; Kulsrud, R.M.

1985-09-01

In high β stellarator and tokamak reactors, the plasma pressure gradient in some regions of the plasma may exceed the critical pressure gradient set by ballooning instabilities. In these regions, convective cells break out to enhance the transport. As a result, the pressure gradient can rise only slightly above the critical gradient and the plasma is in another state of equilibrium - ''convective equilibrium'' - in these regions. Although the convective transport cannot be calculated precisely, it is shown that the density and temperature profiles in the convective region can still be estimated. A simple mixing-length theory, similar to that used for convection in stellar interiors, is introduced in this paper to provide a qualitative description of the convective cells and to show that the convective transport is highly efficient. A numerical example for obtaining the density and temperature profiles in a stellarator reactor is given

In-service ultrasonic inspection of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Prepechal, J.; Sulc, J.

1982-01-01

Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

Acceleration of relativistic electrons in plasma reactors and non-linear spectra of cosmic radio sources

International Nuclear Information System (INIS)

Kaplan, S.A.; Lomadze, R.D.

1978-01-01

A second approximation to the theory of turbulent plasma reactors in connection with the problem of interpretation of the non-linear spectra of cosmic radio sources has been investigated by the authors (Kaplan and Lomadze, 1977; Lomadze, 1977). The present paper discusses the basic results received for a Compton reactor with plasma waves of phase velocities smaller than the velocity of light, as well as for the synchrotron reactor. The distortion of the distribution function of relativistic electrons caused by their diffusion from the reactor is also presented as an example. (Auth.)

Relation between plasma plume density and gas flow velocity in atmospheric pressure plasma

International Nuclear Information System (INIS)

Yambe, Kiyoyuki; Taka, Shogo; Ogura, Kazuo

2014-01-01

We have studied atmospheric pressure plasma generated using a quartz tube, helium gas, and copper foil electrode by applying RF high voltage. The atmospheric pressure plasma in the form of a bullet is released as a plume into the atmosphere. To study the properties of the plasma plume, the plasma plume current is estimated from the difference in currents on the circuit, and the drift velocity is measured using a photodetector. The relation of the plasma plume density n plu , which is estimated from the current and the drift velocity, and the gas flow velocity v gas is examined. It is found that the dependence of the density on the gas flow velocity has relations of n plu ∝ log(v gas ). However, the plasma plume density in the laminar flow is higher than that in the turbulent flow. Consequently, in the laminar flow, the density increases with increasing the gas flow velocity

Voltage uniformity study in large-area reactors for RF plasma deposition

Energy Technology Data Exchange (ETDEWEB)

Sansonnens, L.; Pletzer, A.; Magni, D.; Howling, A.A.; Hollenstein, C. [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP); Schmitt, J.P.M. [Balzers Process Systems, Palaiseau (France)

1996-09-01

Non-uniform voltage distribution across the electrode area results in inhomogeneous thin-film RF plasma deposition in large area reactors. In this work, a two-dimensional analytic model for the calculation of the voltage distribution across the electrode area is presented. The results of this model are in good agreement with measurements performed without plasma at 13.56 MHz and 70 MHz in a large area reactor. The principal voltage inhomogeneities are caused by logarithmic singularities in the vicinity of RF connections and not by standing waves. These singularities are only described by a two-dimensional model and cannot be intuitively predicted by analogy to a one-dimensional case. Plasma light emission measurements and thickness homogeneity studies of a-Si:H films show that the plasma reproduces these voltage inhomogeneities. Improvement of the voltage uniformity is investigated by changing the number and position of the RF connections. (author) 13 figs., 20 refs.

Study of hydroxylation of benzene and toluene using a micro-DBD plasma reactor

International Nuclear Information System (INIS)

Sekiguchi, H; Ando, M; Kojima, H

2005-01-01

The hydroxylation behaviour of benzene and toluene were studied using a micro-plasma reactor, where an atmospheric non-thermal plasma was generated by a dielectric barrier discharge (DBD). The results indicated that oxidation products primarily consisted of phenol and C 4 -compounds for benzene hydroxylation, whereas cresol, benzaldehyde, benzylalcohol and C 4 -compounds were detected for toluene hydroxylation. By taking into consideration the reaction mechanism in the plasma reactor, these products were classified into (1) oxidation of the aromatic ring and functional group on the ring and (2) cleavage of the aromatic ring or dissociation of the functional group on the ring

LOFA analyses for the water and helium cooled SEAFP reactors

International Nuclear Information System (INIS)

Sponton, L.; Sjoeberg, A.; Nordlinder, S.

2001-01-01

This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

Atmospheric-pressure plasma decontamination/sterilization chamber

Science.gov (United States)

Herrmann, Hans W.; Selwyn, Gary S.

2001-01-01

An atmospheric-pressure plasma decontamination/sterilization chamber is described. The apparatus is useful for decontaminating sensitive equipment and materials, such as electronics, optics and national treasures, which have been contaminated with chemical and/or biological warfare agents, such as anthrax, mustard blistering agent, VX nerve gas, and the like. There is currently no acceptable procedure for decontaminating such equipment. The apparatus may also be used for sterilization in the medical and food industries. Items to be decontaminated or sterilized are supported inside the chamber. Reactive gases containing atomic and metastable oxygen species are generated by an atmospheric-pressure plasma discharge in a He/O.sub.2 mixture and directed into the region of these items resulting in chemical reaction between the reactive species and organic substances. This reaction typically kills and/or neutralizes the contamination without damaging most equipment and materials. The plasma gases are recirculated through a closed-loop system to minimize the loss of helium and the possibility of escape of aerosolized harmful substances.

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

Science.gov (United States)

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.

A model for structural analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A. de.

1987-01-01

Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL

2010-08-01

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 regarding 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.

Enriched uranium cycles in pressurized heavy water reactors

International Nuclear Information System (INIS)

Mazzola, A.

1994-01-01

A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

Nuclear power - replacement of pressure tubes in CANDU reactors

International Nuclear Information System (INIS)

Anon.

1992-01-01

The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

International Nuclear Information System (INIS)

Schene, R.

2009-01-01

Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

Non-thermal atmospheric-pressure plasma possible application in wound healing.

Science.gov (United States)

Haertel, Beate; von Woedtke, Thomas; Weltmann, Klaus-Dieter; Lindequist, Ulrike

2014-11-01

Non-thermal atmospheric-pressure plasma, also named cold plasma, is defined as a partly ionized gas. Therefore, it cannot be equated with plasma from blood; it is not biological in nature. Non-thermal atmospheric-pressure plasma is a new innovative approach in medicine not only for the treatment of wounds, but with a wide-range of other applications, as e.g. topical treatment of other skin diseases with microbial involvement or treatment of cancer diseases. This review emphasizes plasma effects on wound healing. Non-thermal atmospheric-pressure plasma can support wound healing by its antiseptic effects, by stimulation of proliferation and migration of wound relating skin cells, by activation or inhibition of integrin receptors on the cell surface or by its pro-angiogenic effect. We summarize the effects of plasma on eukaryotic cells, especially on keratinocytes in terms of viability, proliferation, DNA, adhesion molecules and angiogenesis together with the role of reactive oxygen species and other components of plasma. The outcome of first clinical trials regarding wound healing is pointed out.

EDF's nuclear safety approach for pressurized water reactors

International Nuclear Information System (INIS)

Tanguy, P.; Kus, J.P.

1987-01-01

The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

EDF'S nuclear safety approach for pressurized water reactors

International Nuclear Information System (INIS)

Tanguy, P.; Kus, J.P.

1988-01-01

The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

International Nuclear Information System (INIS)

Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

2001-01-01

The major increase in discharge duration and plasma energy in a next-step DT (deuterium-tritium) fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D (Research and Development) avenues for their resolution are presented

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

Energy Technology Data Exchange (ETDEWEB)

Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

International Nuclear Information System (INIS)

Es-Saheb, M.H.H.

1990-01-01

The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR [pressurized-water-reactor] plants

International Nuclear Information System (INIS)

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1988-01-01

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs

Some considerations on a plasma in the JAERI experimental fusion reactor

International Nuclear Information System (INIS)

Tone, T.; Yamato, H.; Maki, K.

1976-01-01

The preliminary analysis of the plasma characteristics for the JAERI tokamak experimental fusion reactor is reported. In order to make the reactor compact, the self-sustaining condition has been removed. Stationary heating by 200 keV neutral deuteron beam to maintain the power balance is applied expecting the power amplification by the TCT effect. The main parameters determined are power output of 100 MW, toroidal field on axis of 6 T, aspect ratio of 4.5 and major radius of 6.75 m. The results of the plasma power balance, fueling by means of the gas blanket scheme, power stabilization with feedback and the start-up are presented

Plasma pressure and anisotropy inferred from the Tsyganenkomagnetic field model

Directory of Open Access Journals (Sweden)

F. Cao

Full Text Available A numerical procedure has been developed to deduce the plasma pressure and anisotropy from the Tsyganenko magnetic field model. The Tsyganenko empirical field model, which is based on vast satellite field data, provides a realistic description of magnetic field configuration in the magnetosphere. When the force balance under the static condition is assumed, the electromagnetic J×B force from the Tsyganenko field model can be used to infer the plasma pressure and anisotropy distributions consistent with the field model. It is found that the J×B force obtained from the Tsyganenko field model is not curl-free. The curl-free part of the J×B force in an empirical field model can be balanced by the gradient of the isotropic pressure, while the nonzero curl of the J×B force can only be associated with the pressure anisotropy. The plasma pressure and anisotropy in the near-Earth plasma sheet are numerically calculated to obtain a static equilibrium consistent with the Tsyganenko field model both in the noon-midnight meridian and in the equatorial plane. The plasma pressure distribution deduced from the Tsyganenko 1989 field model is highly anisotropic and shows this feature early in the substorm growth phase. The pressure anisotropy parameter α_P, defined as α_P=1-P_VertP_⊥, is typically ~0.3 at x ≈ -4.5R_E and gradually decreases to a small negative value with an increasing tailward distance. The pressure anisotropy from the Tsyganenko 1989 model accounts for 50% of the cross-tail current at maximum and only in a highly localized region near xsim-10R_E. In comparison, the plasma pressure anisotropy inferred from the Tsyganenko 1987 model is much smaller. We also find that the boundary

Effects of Weight Hourly Space Velocity and Catalyst Diameter on Performance of Hybrid Catalytic-Plasma Reactor for Biodiesel Synthesis over Sulphated Zinc Oxide Acid Catalyst

Directory of Open Access Journals (Sweden)

Luqman Buchori

2017-05-01

Full Text Available Biodiesel synthesis through transesterification of soybean oil with methanol on hybrid catalytic-plasma reactor over sulphated zinc oxide (SO42-/ZnO active acid catalyst was investigated. This research was aimed to study effects of Weight Hourly Space Velocity (WHSV and the catalyst diameter on performance of the hybrid catalytic-plasma reactor for biodiesel synthesis. The amount (20.2 g of active sulphated zinc oxide solid acid catalysts was loaded into discharge zone of the reactor. The WHSV and the catalyst diameter were varied between 0.89 to 1.55 min-1 and 3, 5, and 7 mm, respectively. The molar ratio of methanol to oil as reactants of 15:1 is fed to the reactor, while operating condition of the reactor was kept at reaction temperature of 65 oC and ambient pressure. The fatty acid methyl ester (FAME component in biodiesel product was identified by Gas Chromatography - Mass Spectrometry (GC-MS. The results showed that the FAME yield decreases with increasing WHSV. It was found that the optimum FAME yield was achieved of 56.91 % at WHSV of 0.89 min-1 and catalyst diameter of 5 mm and reaction time of 1.25 min. It can be concluded that the biodiesel synthesis using the hybrid catalytic-plasma reactor system exhibited promising the FAME yield. Copyright © 2017 BCREC Group. All rights reserved Received: 15th November 2016; Revised: 24th December 2016; Accepted: 16th February 2017 How to Cite: Buchori, L., Istadi, I., Purwanto, P. (2017. Effects of Weight Hourly Space Velocity and Catalyst Diameter on Performance of Hybrid Catalytic-Plasma Reactor for Biodiesel Synthesis over Sulphated Zinc Oxide Acid Catalyst. Bulletin of Chemical Reaction Engineering & Catalysis, 12 (2: 227-234 (doi:10.9767/bcrec.12.2.775.227-234 Permalink/DOI: http://dx.doi.org/10.9767/bcrec.12.2.775.227-234

Pressure suppression system for a nuclear reactor

International Nuclear Information System (INIS)

Jost, N.

1977-01-01

The invention pertains to a pressure suppression system for PWR reactors where the parts enclosing the primary coolant are contained in two pressure-tight separate chambers. According to the invention, these chambers are partly filled with water and are connected with each other below the water surface. This way, gases cannot escape from the containment, not even if a valve and a line are damaged at the same time, as the vapours released condensate in the water of at least one of the other chambers. (HP) [de

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1978-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1977-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

Stresses in reactor pressure vessel nozzles -- Calculations and experiments

International Nuclear Information System (INIS)

Brumovsky, M.; Polachova, H.

1995-01-01

Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

Research on plasma core reactors

International Nuclear Information System (INIS)

Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

1977-01-01

Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

Nuclear reactor installation with outer shell enclosing a primary pressure vessel

International Nuclear Information System (INIS)

1975-01-01

The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

International Nuclear Information System (INIS)

Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

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-temperature (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

Shapes of agglomerates in plasma etching reactors

International Nuclear Information System (INIS)

Huang, F.Y.; Kushner, M.J.

1997-01-01

Dust particle contamination of wafers in reactive ion etching (RIE) plasma tools is a continuing concern in the microelectronics industry. It is common to find that particles collected on surfaces or downstream of the etch chamber are agglomerates of smaller monodisperse spherical particles. The shapes of the agglomerates vary from compact, high fractal dimension structures to filamentary, low fractal dimension structures. These shapes are important with respect to the transport of particles in RIE tools under the influence electrostatic and ion drag forces, and the possible generation of polarization forces. A molecular dynamics simulation has been developed to investigate the shapes of agglomerates in plasma etching reactors. We find that filamentary, low fractal dimension structures are generally produced by smaller (<100s nm) particles in low powered plasmas where the kinetic energy of primary particles is insufficient to overcome the larger Coulomb repulsion of a compact agglomerate. This is analogous to the diffusive regime in neutral agglomeration. Large particles in high powered plasmas generally produce compact agglomerates of high fractal dimension, analogous to ballistic agglomeration of neutrals. copyright 1997 American Institute of Physics

Dual pressurized light water reactor producing 2000 M We

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

Activities in the Czech Republic for reactor pressure components lifetime management

International Nuclear Information System (INIS)

Brumovsky, M.

1994-01-01

The following activities in the Czech republic for reactor pressure components lifetime management are described: upgrading and safety assurance of nuclear power plants (NPP) with reactors of WWER-440/V-230 type, safety assurance of NPPs with reactors of WWER-440/V-213, lifetime management programme of NPPs with WWER-440/V-213 reactors, preparation of start-up of NPPs with WWER-1000/V-320 reactors, preparation of guides for lifetime as well as defect allowability evaluation in main components of primary and secondary circuits. 3 figs

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

Science.gov (United States)

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.

Preventive protection device and method for bottom of reactor pressure vessel

International Nuclear Information System (INIS)

Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

1995-01-01

In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

Co-generation of synthesis gas and C{sub 2+} hydrocarbons from methane and carbon dioxide in a hybrid catalytic-plasma reactor: A review

Energy Technology Data Exchange (ETDEWEB)

Istadi; Nor Aishah Saidina Amin [Universiti Teknologi Malaysia, Johor Bahru (Malaysia). Chemical Reaction Engineering Group (CREG), Faculty of Chemical and Natural Resources Engineering

2006-03-15

The topics on conversion and utilization of methane and carbon dioxide are important issues in tackling the global warming effects from the two greenhouse gases. Several technologies including catalytic and plasma have been proposed to improve the process involving conversion and utilization of methane and carbon dioxide. In this paper, an overview of the basic principles, and the effects of CH{sub 4}/CO{sub 2} feed ratio, total feed flow rate, discharge power, catalyst, applied voltage, wall temperature, and system pressure in dielectric-barrier discharge (DBD) plasma reactor are addressed. The discharge power, discharge gap, applied voltage and CH{sub 4}/CO{sub 2} ratio in the feed showed the most significant effects on the reactor performance. Co-feeding carbon dioxide with the methane feed stream reduced coking and increased methane conversion. The H{sub 2}/CO ratio in the products was significantly affected by CH{sub 4}/CO{sub 2} ratio. The synergism of the catalyst placed in the discharge gap and the plasma affected the products distribution significantly. Methane and carbon dioxide conversions were influenced significantly by discharge power and applied voltage. The drawbacks of DBD plasma application in the CH{sub 4}-CO{sub 2} conversion should be taken into consideration before a new plausible reactor system can be implemented. 76 refs., 4 figs., 2 tabs.

Titanium nitride plasma-chemical synthesis with titanium tetrachloride raw material in the DC plasma-arc reactor

Science.gov (United States)

Kirpichev, D. E.; Sinaiskiy, M. A.; Samokhin, A. V.; Alexeev, N. V.

2017-04-01

The possibility of plasmochemical synthesis of titanium nitride is demonstrated in the paper. Results of the thermodynamic analysis of TiCl4 - H2 - N2 system are presented; key parameters of TiN synthesis process are calculated. The influence of parameters of plasma-chemical titanium nitride synthesis process in the reactor with an arc plasmatron on characteristics on the produced powders is experimentally investigated. Structure, chemical composition and morphology dependencies on plasma jet enthalpy, stoichiometric excess of hydrogen and nitrogen in a plasma jet are determined.

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

Highly Radiative Plasmas for Local Transport Studies and Power and Particle Handling in Reactor Regimes

International Nuclear Information System (INIS)

Bell, M.G.; Bell, R.E.; Budny, R.; Bush, C.E.; Hill, K.W.

1998-01-01

To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into the Tokamak Fusion Test Reactor (TFTR) supershots and high-l(subscript) plasmas. At neutral beam injection (NBI) powers P(subscript B) greater than or equal to 30 MW, carbon influxes (blooms) were suppressed, leading to improved energy confinement and neutron production in both deuterium (D) and deuterium-tritium (DT) plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly-radiative plasmas in the International Thermonuclear Experimental Reactor (ITER). The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms

Methodology of fuel rod design for pressurized light water reactors

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Esteves, A.M.

1988-01-01

The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

Hydrophilic surface modification of coronary stent using an atmospheric pressure plasma jet for endothelialization.

Science.gov (United States)

Shim, Jae Won; Bae, In-Ho; Park, Dae Sung; Lee, So-Youn; Jang, Eun-Jae; Lim, Kyung-Seob; Park, Jun-Kyu; Kim, Ju Han; Jeong, Myung Ho

2018-03-01

The first two authors contributed equally to this study. Bioactivity and cell adhesion properties are major factors for fabricating medical devices such as coronary stents. The aim of this study was to evaluate the advantages of atmospheric-pressure plasma jet in enhancing the biocompatibility and endothelial cell-favorites. The experimental objects were divided into before and after atmospheric-pressure plasma jet treatment with the ratio of nitrogen:argon = 3:1, which is similar to air. The treated surfaces were basically characterized by means of a contact angle analyzer for the activation property on their surfaces. The effect of atmospheric-pressure plasma jet on cellular response was examined by endothelial cell adhesion and XTT analysis. It was difficult to detect any changeable morphology after atmospheric-pressure plasma jet treatment on the surface. The roughness was increased after atmospheric-pressure plasma jet treatment compared to nonatmospheric-pressure plasma jet treatment (86.781 and 7.964 nm, respectively). The X-ray photoelectron spectroscopy results showed that the surface concentration of the C-O groups increased slightly from 6% to 8% after plasma activation. The contact angle dramatically decreased in the atmospheric-pressure plasma jet treated group (22.6 ± 15.26°) compared to the nonatmospheric-pressure plasma jet treated group (72.4 ± 15.26°) ( n = 10, p atmospheric-pressure plasma jet on endothelial cell migration and proliferation was 85.2% ± 12.01% and 34.2% ± 2.68%, respectively, at 7 days, compared to the nonatmospheric-pressure plasma jet treated group (58.2% ± 11.44% in migration, n = 10, p atmospheric-pressure plasma jet method. Moreover, the atmospheric-pressure plasma jet might affect re-endothelialization after stenting.

Improvement of Algorithms for Pressure Maintenance Systems in Drum-Separators of RBMK-1000 Reactors

Energy Technology Data Exchange (ETDEWEB)

Aleksakov, A. N., E-mail: yankovskiy.k@nikiet.ru; Yankovskiy, K. I. [JSC “N. A. Dollezhal Research and Development Institute of Power Engineering (NIKIET),” (Russian Federation); Dunaev, V. I.; Kushbasov, A. N. [JSC “Diakont,” (Russian Federation)

2015-05-15

The main tasks and challenges for pressure regulation in the drum-separators of RBMK-1000 reactors are described. New approaches to constructing algorithms for pressure control in drum-separators by electro-hydraulic turbine control systems are discussed. Results are provided from tests of the operation of modernized pressure regulators during fast transients with reductions in reactor power.

Dual shell pressure balanced reactor vessel. Final project report

International Nuclear Information System (INIS)

Robertus, R.J.; Fassbender, A.G.

1994-10-01

The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

Effects of gas-flow structures on radical and etch-product density distributions on wafers in magnetomicrowave plasma etching reactors

International Nuclear Information System (INIS)

Ikegawa, Masato; Kobayashi, Jun'ichi; Fukuyama, Ryoji

2001-01-01

To achieve high etch rate, uniformity, good selectivity, and etch profile control across large diameter wafers, the distributions of ions, radicals, and etch products in magnetomicrowave high-etch-rate plasma etching reactors must be accurately controlled. In this work the effects of chamber heights, a focus ring around the wafer, and gas supply structures (or gas flow structures) on the radicals and etch products flux distribution onto the wafer were examined using the direct simulation Monte Carlo method and used to determine the optimal reactor geometry. The pressure uniformity on the wafer was less than ±1% when the chamber height was taller than 60 mm. The focus ring around the wafer produced uniform radical and etch-product fluxes but increased the etch-product flux on the wafer. A downward-flow gas-supply structure (type II) produced a more uniform radical distribution than that produced by a radial gas-supply structure (type I). The impact flow of the type II structure removed etch products from the wafer effectively and produced a uniform etch-product distribution even without the focus ring. Thus the downward-flow gas-supply structure (type II) was adopted in the design for the second-generation of a magnetomicrowave plasma etching reactor with a higher etching rate

Calculation of high-pressure argon plasma parameters produced by excimer laser

International Nuclear Information System (INIS)

Tsuda, Norio; Yamada, Jun

2000-01-01

When a XeCl excimer laser light was focused in a high-pressure argon gas up to 150 atm, a dense plasma developed not only backward but also forward. It is important to study on the electron density and temperature of the laser-induced plasma in the high-pressure gas. The electron density and temperature in high-pressure argon plasma produced by XeCl excimer laser has been calculated and compared with the experimental data. (author)

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Dense Medium Plasma Water Purification Reactor (DMP WaPR), Phase I

Data.gov (United States)

National Aeronautics and Space Administration — The Dense Medium Plasma Water Purification Reactor offers significant improvements over existing water purification technologies used in Advanced Life Support...

Basic principles and applications of atmospheric-pressure discharge plasmas

International Nuclear Information System (INIS)

Becker, K.H.

2002-01-01

The principles that govern the generation and maintenance of atmospheric - pressure discharge plasmas are summarized. The properties and operating parameters of various types such as dielectric barrier discharge plasmas (DBDs), corona discharge plasmas (CDs), microhollow cathode discharge plasmas (MHCDs) , and dielectric capillary electrode discharge plasmas (CDEDs) are introduced. All of them are self sustained, non equilibrium gas discharges that can be operated at atmospheric pressure. CDs and DBDDs represent very similar types of discharges, while DBDs are characterized by insulating layers on one or both electrodes, CDs depend on inhomogeneous electric fields at least in some parts of the electrode configuration to restrict the primary ionization processes to a small fraction of the inter - electrode region. Their application to novel light sources in the ultraviolet (UV) and vacuum ultraviolet (VUV) spectral region is described. (nevyjel)

Performance of pressure tubes in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

Influence of Plasma Pressure Fluctuation on RF Wave Propagation

International Nuclear Information System (INIS)

Liu Zhiwei; Bao Weimin; Li Xiaoping; Liu Donglin; Zhou Hui

2016-01-01

Pressure fluctuations in the plasma sheath from spacecraft reentry affect radio-frequency (RF) wave propagation. The influence of these fluctuations on wave propagation and wave properties is studied using methods derived by synthesizing the compressible turbulent flow theory, plasma theory, and electromagnetic wave theory. We study these influences on wave propagation at GPS and Ka frequencies during typical reentry by adopting stratified modeling. We analyzed the variations in reflection and transmission properties induced by pressure fluctuations. Our results show that, at the GPS frequency, if the waves are not totally reflected then the pressure fluctuations can remarkably affect reflection, transmission, and absorption properties. In extreme situations, the fluctuations can even cause blackout. At the Ka frequency, the influences are obvious when the waves are not totally transmitted. The influences are more pronounced at the GPS frequency than at the Ka frequency. This suggests that the latter can mitigate blackout by reducing both the reflection and the absorption of waves, as well as the influences of plasma fluctuations on wave propagation. Given that communication links with the reentry vehicles are susceptible to plasma pressure fluctuations, the influences on link budgets should be taken into consideration. (paper)

A dual pressurized water reactor producing 2000 MWe

International Nuclear Information System (INIS)

Kang, K. M.; Suh, K. Y.

2010-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

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

Science.gov (United States)

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.

A Study on Decontamination Process Using Atmospheric Pressure Plasma

International Nuclear Information System (INIS)

Kim, Yong Soo; Jeon, Sang Hwan; Jin, Dong Sik; Park, Dong Min

2010-05-01

Radioactive decontamination process using atmospheric pressure plasma which can be operated parallel with low vacuum cold plasma processing is studied. Two types of cold plasma torches were designed and manufactured. One of them is the cylindrical type applicable to the treatment of three-dimensional surfaces. The other is the rectangular type for the treatment of flat and large surface areas. Ar palsam was unstable but using He as a carrier gas, discharge condition was improved. Besides filtering module using pre, medium, charcoal, and HEPA filter was designed and manufactured. More intensive study for developing filtering system will be followed. Atmospheric pressure plasma decontamination process can be used to the equipment and facility wall decontamination

Analysis of a gas-liquid film plasma reactor for organic compound oxidation

Energy Technology Data Exchange (ETDEWEB)

Hsieh, Kevin [Department of Chemical and Biomedical Engineering, Florida State University, Tallahassee, FL 32310 (United States); Wang, Huijuan [School of Environmental and Safety Engineering, Jiangsu University, Zhenjiang 212013 (China); Locke, Bruce R., E-mail: blocke@fsu.edu [Department of Chemical and Biomedical Engineering, Florida State University, Tallahassee, FL 32310 (United States)

2016-11-05

Highlights: • Non-homogeneous filamentary plasma discharge formed along gas-liquid interface. • Hydrogen peroxide formed near interface favored over organic oxidation from liquid. • Post-plasma Fenton reactions lead to complete utilization of hydrogen peroxide. - Abstract: A pulsed electrical discharge plasma formed in a tubular reactor with flowing argon carrier gas and a liquid water film was analyzed using methylene blue as a liquid phase hydroxyl radical scavenger and simultaneous measurements of hydrogen peroxide formation. The effects of liquid flow rate, liquid conductivity, concentration of dye, and the addition of ferrous ion on dye decoloration and degradation were determined. Higher liquid flow rates and concentrations of dye resulted in less decoloration percentages and hydrogen peroxide formation due to initial liquid conductivity effects and lower residence times in the reactor. The highest decoloration energy yield of dye found in these studies was 5.2 g/kWh when using the higher liquid flow rate and adding the catalyst. The non-homogeneous nature of the plasma discharge favors the production of hydrogen peroxide in the plasma-liquid interface over the chemical oxidation of the organic in the bulk liquid phase and post-plasma reactions with the Fenton catalyst lead to complete utilization of the plasma-formed hydrogen peroxide.

A global model for SF6 plasmas coupling reaction kinetics in the gas phase and on the surface of the reactor walls

International Nuclear Information System (INIS)

Kokkoris, George; Panagiotopoulos, Apostolos; Gogolides, Evangelos; Goodyear, Andy; Cooke, Mike

2009-01-01

Gas phase and reactor wall-surface kinetics are coupled in a global model for SF 6 plasmas. A complete set of gas phase and surface reactions is formulated. The rate coefficients of the electron impact reactions are based on pertinent cross section data from the literature, which are integrated over a Druyvesteyn electron energy distribution function. The rate coefficients of the surface reactions are adjustable parameters and are calculated by fitting the model to experimental data from an inductively coupled plasma reactor, i.e. F atom density and pressure change after the ignition of the discharge. The model predicts that SF 6 , F, F 2 and SF 4 are the dominant neutral species while SF 5 + and F - are the dominant ions. The fit sheds light on the interaction between the gas phase and the reactor walls. A loss mechanism for SF x radicals by deposition of a fluoro-sulfur film on the reactor walls is needed to predict the experimental data. It is found that there is a net production of SF 5 , F 2 and SF 6 , and a net consumption of F, SF 3 and SF 4 on the reactor walls. Surface reactions as well as reactions between neutral species in the gas phase are found to be important sources and sinks of the neutral species.

The stabilizing effect of core pressure on the edge pedestal in MAST plasmas

International Nuclear Information System (INIS)

Chapman, I.T.; Simpson, J.; Saarelma, S.; Kirk, A.; O'Gorman, T.; Scannell, R.

2015-01-01

The pedestal pressure measured in Mega Ampere Spherical Tokamak plasmas has been shown to increase as the global plasma pressure increases. By deliberately suppressing the transition into the high-confinement regime, the core plasma pressure was systematically altered at the time of the first edge localized mode. Stability analysis shows that the enhanced Shafranov shift at higher core pressure stabilizes the ballooning modes driven by the pedestal pressure gradient, consequently allowing the pedestal to reach higher pressures. (paper)

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Najzer, M.

1982-01-01

Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

Startup of reversed-field mirror reactors using coaxial plasma guns

International Nuclear Information System (INIS)

Smith, A.C. Jr.; Hartman, C.W.; Carlson, G.A.; Neef, W.S. Jr.; Eddleman, J.L.

1979-01-01

Preliminary calculations are given that indicate that a coaxial plasma gun might scale reasonably to reactor-grade operating conditions. Ongoing experiments and numerical simulations should shed some light on the validity of the described scaling laws

In-service inspection program for the NCS-80 reactor pressure vessel

International Nuclear Information System (INIS)

Scharge, J.; Wehowsky, P.; Zeibig, H.

1978-01-01

The in-service inspection program of reactor pressure vessels is mainly based on the ultra-sonic method, visual checking of inner and outer surfaces as well as pressure and leak tests. The test procedure require a design of the pressure vessel suitable for the test methods and the possibility to remove the pressure vessel internals. For the outside inspection a gap of sufficient width is mandatory. The present status of the ultra-sonic method and of the inner and outer manipulators affords to conduct the in-service inspection program in form of automatic checkings. The in-service inspection program for NCS-80, the Nuclear Container-Ship design of 80,000 shp, is integrated in the refueling periods due to the request for a high availability of the ship and reactor plant

Diagnostics of plasma-biological surface interactions in low pressure and atmospheric pressure plasmas

International Nuclear Information System (INIS)

Ishikawa, Kenji; Hori, Masaru

2014-01-01

Mechanisms of plasma-surface interaction are required to understand in order to control the reactions precisely. Recent progress in atmospheric pressure plasma provides to apply as a tool of sterilization of contaminated foodstuffs. To use the plasma with safety and optimization, the real time in situ detection of free radicals - in particular dangling bonds by using the electron-spin-resonance (ESR) technique has been developed because the free radical plays important roles for dominantly biological reactions. First, the kinetic analysis of free radicals on biological specimens such as fungal spores of Penicillium digitatum interacted with atomic oxygen generated plasma electric discharge. We have obtained information that the in situ real time ESR signal from the spores was observed and assignable to semiquinone radical with a g-value of around 2.004 and a line width of approximately 5G. The decay of the signal was correlated with a link to the inactivation of the fungal spore. Second, we have studied to detect chemical modification of edible meat after the irradiation. Using matrix-assisted laser desorption/ionization time-of-flight mass spectroscopy (MALDI-TOF-MS) and ESR, signals give qualification results for chemical changes on edible liver meat. The in situ real-time measurements have proven to be a useful method to elucidate plasma-induced surface reactions on biological specimens. (author)

Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

International Nuclear Information System (INIS)

Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

1994-01-01

This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

Light pressure of time-dependent fields in plasmas

International Nuclear Information System (INIS)

Zeidler, A.; Schnabl, H.; Mulser, P.

1985-01-01

An expression of the light pressure Pi is derived for the case of a nearly monochromatic electromagnetic wave with arbitrarily time-dependent amplitude. Thereby Pi is defined as the time-averaged force density exerted on a plasma by the wave. The resulting equations are valid for both transverse and longitudinal waves. The light pressure turns out to consist of two components: the well-known gradient-type term and a new nonstationary solenoidal term. This is true for warm as well as cold plasmas. The importance of the new term for the generation of static magnetic fields is shown, and a model in which shear forces may result is given. Formulas for the nonstationary light pressure developed previously are discussed

Influence of decontamination of the WWER-440 primary circuit equipment on pressure drop in the reactor

International Nuclear Information System (INIS)

Kritsky, V.; Rodionov, Y.; Beresina, I.

2003-01-01

Over 40 reactor cycles at four WWER-440 type reactors have been analyzed in order to explain the increase of the pressure drop under certain combination of conditions. It is shown that the staff radiation exposure and the dose rate at first circuit segments are inversely correlated with the value of the pressure drop at the reactor, which is connected with the mechanism of redistribution of deposits and radioactive nuclides between the reactor and the rest part of the circuit. The influence of pH T on the formation of the dose rate from equipment and the change of pressure drop in the reactor WWER-440 is studied. The optimal range of pH T values for these parameters is determined to be 6.95-7.05 and these values are within the range of the water chemistry standards. The correlation between the changes of pressure drop and the number of decontaminated steam generators is established. This correlation shows that the pressure drop at the reactor grows with the increase of steam generators decontaminated during a preventive maintenance

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

Non-equilibrium plasma chemistry at high pressure and its applications

International Nuclear Information System (INIS)

Bai Xiyao; Zhang Zhitao; Bai Mindong; Zhu Qiaoying

2000-01-01

A review is presented of research and development of gas discharge and non-equilibrium plasma including, new ideas of non-equilibrium plasma at high gas pressure. With special technology, strong electric fields (>400 Td) can be achieved by which electrons are accelerated suddenly, becoming high energy electrons (> 10 eV) at high pressure. On impact with the electrons, the gas molecules dissociate into ions, atomic ions, atoms and free radicals, and new substances or molecules can be synthesized through custom design. Chemical reaction difficult to achieve by conventional method can be realized or accelerated. Non-equilibrium plasma chemistry at high pressure has wide application prospects

Development of dispersion interferometer for magnetic confinement plasmas and high-pressure plasmas

Science.gov (United States)

Akiyama, T.; Yasuhara, R.; Kawahata, K.; Nakayama, K.; Okajima, S.; Urabe, K.; Terashima, K.; Shirai, N.

2015-09-01

A CO2 laser dispersion interferometer (DI) has been developed for both magnetically fusion plasmas and high pressure industrial plasmas. The DI measures the phase shift caused by dispersion in a medium. Therefore, it is insensitive to the mechanical vibrations and changes in the neutral gas density, which degrade the resolution of the electron density measurement. We installed the DI on the Large Helical Device (LHD) and demonstrated a high density resolution of 2× 1017 m-3 without any vibration-free bench. The measured electron density with the DI shows good agreement with results of the existing far infrared laser (a wavelength of 119 μ m) interferometer. The DI system is also applied to the electron density measurement of high-pressure small-scale plasmas. The significant suppression of the phase shift caused by the neutral gas is proven. The achieved density resolution was 1.5× 1019 m-3 with a response time of 100 μ s. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

Intracellular effects of atmospheric-pressure plasmas on melanoma cancer cells

Energy Technology Data Exchange (ETDEWEB)

Ishaq, M., E-mail: ishaqmusarat@gmail.com [Peter MacCallum Cancer Centre, East Melbourne, VIC 3002 (Australia); Comonwealth Scientific and Industrial Research Organization, Sydney, New South Wales (Australia); Bazaka, K. [Institute for Health and Biomedical Innovation, School of Chemistry, Physics and Mechanical Engineering, Queensland University of Technology, Brisbane, QLD 4000 (Australia); Ostrikov, K. [Comonwealth Scientific and Industrial Research Organization, Sydney, New South Wales (Australia); Institute for Health and Biomedical Innovation, School of Chemistry, Physics and Mechanical Engineering, Queensland University of Technology, Brisbane, QLD 4000 (Australia)

2015-12-15

Gas discharge plasmas formed at atmospheric pressure and near room temperature have recently been shown as a promising tool for cancer treatment. The mechanism of the plasma action is attributed to generation of reactive oxygen and nitrogen species, electric fields, charges, and photons. The relative importance of different modes of action of atmospheric-pressure plasmas depends on the process parameters and specific treatment objects. Hence, an in-depth understanding of biological mechanisms that underpin plasma-induced death in cancer cells is required to optimise plasma processing conditions. Here, the intracellular factors involved in the observed anti-cancer activity in melanoma Mel007 cells are studied, focusing on the effect of the plasma treatment dose on the expression of tumour suppressor protein TP73. Over-expression of TP73 causes cell growth arrest and/or apoptosis, and hence can potentially be targeted to enhance killing efficacy and selectivity of the plasma treatment. It is shown that the plasma treatment induces dose-dependent up-regulation of TP73 gene expression, resulting in significantly elevated levels of TP73 RNA and protein in plasma-treated melanoma cells. Silencing of TP73 expression by means of RNA interference inhibited the anticancer effects of the plasma, similar to the effect of caspase inhibitor z-VAD or ROS scavenger N-acetyl cysteine. These results confirm the role of TP73 protein in dose-dependent regulation of anticancer activity of atmospheric-pressure plasmas.

Hydrophilic property of 316L stainless steel after treatment by atmospheric pressure corona streamer plasma using surface-sensitive analyses

Energy Technology Data Exchange (ETDEWEB)

Al-Hamarneh, Ibrahim, E-mail: hamarnehibrahim@yahoo.com [Department of Physics, Faculty of Science, Al-Balqa Applied University, Salt 19117 (Jordan); Pedrow, Patrick [School of Electrical Engineering and Computer Science, Washington State University, Pullman, WA 99164 (United States); Eskhan, Asma; Abu-Lail, Nehal [Gene and Linda Voiland School of Chemical Engineering and Bioengineering, Washington State University, Pullman, WA 99164 (United States)

2012-10-15

Highlights: Black-Right-Pointing-Pointer Surface hydrophilic property of surgical-grade 316L stainless steel was enhanced by Ar-O{sub 2} corona streamer plasma treatment. Black-Right-Pointing-Pointer Hydrophilicity, surface morphology, roughness, and chemical composition before and after plasma treatment were evaluated. Black-Right-Pointing-Pointer Contact angle measurements and surface-sensitive analyses techniques, including XPS and AFM, were carried out. Black-Right-Pointing-Pointer Optimum plasma treatment conditions of the SS 316L surface were determined. - Abstract: Surgical-grade 316L stainless steel (SS 316L) had its surface hydrophilic property enhanced by processing in a corona streamer plasma reactor using O{sub 2} gas mixed with Ar at atmospheric pressure. Reactor excitation was 60 Hz ac high-voltage (0-10 kV{sub RMS}) applied to a multi-needle-to-grounded screen electrode configuration. The treated surface was characterized with a contact angle tester. Surface free energy (SFE) for the treated stainless steel increased measurably compared to the untreated surface. The Ar-O{sub 2} plasma was more effective in enhancing the SFE than Ar-only plasma. Optimum conditions for the plasma treatment system used in this study were obtained. X-ray photoelectron spectroscopy (XPS) characterization of the chemical composition of the treated surfaces confirms the existence of new oxygen-containing functional groups contributing to the change in the hydrophilic nature of the surface. These new functional groups were generated by surface reactions caused by reactive oxidation of substrate species. Atomic force microscopy (AFM) images were generated to investigate morphological and roughness changes on the plasma treated surfaces. The aging effect in air after treatment was also studied.

Plasma acceleration by means of microwave radiation pressure

International Nuclear Information System (INIS)

Fukumura, Takashi; Takamoto, Teruo

1977-01-01

In the electric discharge of gas with microwaves, intense reflection waves occur simultaneously with the discharge, so the plasma ionized and formed by the microwaves is accelerated due to large radiation pressure. The basic experiment made, aiming at plasma gun, is described. In the gas electric discharge, the plasma flow velocity proportional to the reflected power is obtained. For 550 W microwave input power, the plasma velocity of 1 x 10 4 m/s was obtained. The accelerated plasma is bunched; its front as mass travels, recombines and disappears. (Mori, K.)

Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel

International Nuclear Information System (INIS)

Chou, Hsoung-Wei; Huang, Chin-Cheng

2017-01-01

Highlights: • P-T limits based on ASME K_I_a curve, K_I_C curve and RI method are presented. • Probabilistic and deterministic methods are used to evaluate P-T limits on RPV. • The feasibility of substituting P-T curves with more operational is demonstrated. • Warm-prestressing effect is critical in determining the fracture probability. - Abstract: The ASME Code Section XI-Appendix G defines the normal reactor startup (heat-up) and shut-down (cool-down) operation limits according to the fracture toughness requirement of reactor pressure vessel (RPV) materials. This paper investigates the effects of different pressure-temperature limit operations on structural integrity of a Taiwan domestic pressurized water reactor (PWR) pressure vessel. Three kinds of pressure-temperature limits based on different fracture toughness requirements – the K_I_a fracture toughness curve of ASME Section XI-Appendix G before 1998 editions, the K_I_C fracture toughness curve of ASME Section XI-Appendix G after 2001 editions, and the risk-informed revision method supplemented in ASME Section XI-Appendix G after 2013 editions, respectively, are established as the loading conditions. A series of probabilistic fracture mechanics analyses for the RPV are conducted employing ORNL’s FAVOR code considering various radiation embrittlement levels under these pressure-temperature limit conditions. It is found that the pressure-temperature operation limits which provide more operational flexibility may lead to higher fracture risks to the RPV. The cladding-induced shallow surface breaking flaws are the most critical and dominate the fracture probability of the RPV under pressure-temperature limit transients. Present study provides a risk-informed reference for the operation safety and regulation viewpoint of PWRs in Taiwan.

Plasma engineering design of a compact reversed-field pinch reactor (CRFPR)

International Nuclear Information System (INIS)

Bathke, C.G.; Embrechts, M.J.; Hagenson, R.L.; Krakowski, R.A.; Miller, R.L.

1983-01-01

The rationale for and the characteristics of the high-power-density Compact Reversed-Field Pinch Reactor (CRFPR) are discussed. Particular emphasis is given to key plasma engineering aspects of the conceptual design, including plasma operations, current drive, and impurity/ash control by means of pumped limiters or magnetic divertors. A brief description of the Fusion-Power-Core integration is given

Integral Pressurized Water Reactor Simulator Manual

International Nuclear Information System (INIS)

2017-01-01

This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

Very low pressure plasma sprayed yttria-stabilized zirconia coating using a low-energy plasma gun

International Nuclear Information System (INIS)

Zhu, Lin; Zhang, Nannan; Bolot, Rodolphe; Planche, Marie-Pierre; Liao, Hanlin; Coddet, Christian

2011-01-01

In the present study, a more economical low-energy plasma source was used to perform a very low pressure plasma-spray (VLPPS) process. The plasma-jet properties were analyzed by means of optical emission spectroscopy (OES). Moreover, yttria-stabilized zirconia coating (YSZ) was elaborated by a F100 low-power plasma gun under working pressure of 1 mbar, and the substrate specimens were partially shadowed by a baffle-plate during plasma spraying for obtaining different coating microstructures. Based on the SEM observation, a column-like grain coating was deposited by pure vapor deposition at the shadowed region, whereas, in the unshadowed region, the coating exhibited a binary microstructure which was formed by a mixed deposition of melted particles and evaporated particles. The mechanical properties of the coating were also well under investigation. (orig.)

Experiments on the spray nozzles used in the pressurizer of power reactor

International Nuclear Information System (INIS)

Diao Wentang

1989-04-01

The spray nozzle, which is used in the pressurizer of pressurized water reactor system, usually uses a less differential pressure between the reactor inlet and outlet as the spray drive pressure, but its flow rate is relatively larger. It is difficult to obtain a optimum spray performance of such a nozzle. The experimental results of five types of twenty seven spray nozzles in different structures and sizes with the range of the spray drive pressure from 0.127 to 0.245 MPa and the flow rates from 5 to 50 t/h are given. The main factors affecting spray performances and their distribution characteristics have been found. And some relatively suitable spray structures have been recommended, which can be used as references for improving the spray nozzles used in the pressurizers of existing PWRs or of the PWRs to be built

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

Science.gov (United States)

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.

Instrumentation and control strategies for an integral pressurized water reactor

Directory of Open Access Journals (Sweden)

Belle R. Upadhyaya

2015-03-01

Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

Characterization of DC argon plasma jet at atmospheric pressure

International Nuclear Information System (INIS)

Yan Jianhua; Ma Zengyi; Pan Xinchao; Cen Kefa; Bruno, C

2006-01-01

An original DC double anode plasma torch operating with argon at atmospheric pressure which provides a long time and highly stable plasma jet is analyzed through its electrical and optical signals. Effects of gas flow rate and current intensity on the arc dynamics behaviour are studied using standard diagnostic tools such as FFT and correlation function. An increasing current-voltage characteristic is reported for different argon flow rates. It is noted that the takeover mode is characteristic for argon plasma jet and arc fluctuations in our case are mainly induced by the undulation of torch power supply. Furthermore, the excitation temperatures and electron densities of the plasma jet inside and outside the arc chamber have been determined by means of optical emission spectroscopy (OES). The criteria for the existence of local thermodynamic equilibrium (LTE) in plasma is then discussed. The results show that argon plasma jet at atmospheric pressure under our experimental conditions is close to LTE. (authors)

Assessment of the integrity of WWER type reactor pressure vessels

International Nuclear Information System (INIS)

Brumovsky, M.

1995-01-01

Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

Low pressure water vapour plasma treatment of surfaces for biomolecules decontamination

DEFF Research Database (Denmark)

Fumagalli, F; Kylian, O; Amato, Letizia

2012-01-01

Decontamination treatments of surfaces are performed on bacterial spores, albumin and brain homogenate used as models of biological contaminations in a low-pressure, inductively coupled plasma reactor operated with water-vapour-based gas mixtures. It is shown that removal of contamination can...... be achieved using pure H2O or Ar/H2O mixtures at low temperatures with removal rates comparable to oxygen-based mixtures. Particle fluxes (Ar+ ions, O and H atomic radicals and OH molecular radicals) from water vapour discharge are measured by optical emission spectroscopy and Langmuir probe under several...... operating conditions. Analysis of particle fluxes and removal rates measurements illustrates the role of ion bombardment associated with O radicals, governing the removal rates of organic matter. Auxiliary role of hydroxyl radicals is discussed on the basis of experimental data. The advantages of a water...

Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

International Nuclear Information System (INIS)

Virgilio, M.J.

1981-11-01

The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

Self-actuated rate of change of pressure scram device for nuclear reactors

International Nuclear Information System (INIS)

Noyes, R.C.; Zaman, S.U.; Stuteville, D.W.

1979-01-01

A sensor chamber having one cavity containing coolant separated by a diaphragm from another cavity containing a fixed mass of inert gas is located within a safety assembly of a liquid metal-cooled nuclear reactor. The liquid cavity is in fluid communication with the coolant outside the chamber through a flow limiting orifice. An actuating bellows in fluid communication with the gas cavity is in contact with coolant outside the chamber and is connected to a push rod, which serves as a trigger for a poison bundle relase mechanism. During slow changes in reactor coolant pressure experienced under normal operation, the diaphragm moves to equalize the gas cavity and liquid cavity pressures with the coolant pressure outside the chamber. The actuating bellows does not move because it is biased so that a threshold pressure difference is required before it will expand. Under a more rapid drop in coolant pressure, such as is associated with a loss of forced flow, the threshold is overcome and the actuating bellows will also move, thereby triggering the release mechanism to shut down the reactor. The actuating bellows may be connected to the liquid cavity rather than to the gas cavity

A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

International Nuclear Information System (INIS)

Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

1997-01-01

A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

1997-04-01

A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

Science.gov (United States)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were

Atmospheric-pressure guided streamers for liposomal membrane disruption

International Nuclear Information System (INIS)

Svarnas, P.; Aleiferis, Sp.; Matrali, S. H.; Gazeli, K.; Clément, F.; Antimisiaris, S. G.

2012-01-01

The potential to use liposomes (LIPs) as a cellular model in order to study interactions of cold atmospheric-pressure plasma with cells is herein investigated. Cold atmospheric-pressure plasma is formed by a dielectric-barrier discharge reactor. Large multilamellar vesicle liposomes, consisted of phosphatidylcholine and cholesterol, are prepared by the thin film hydration technique, to encapsulate a small hydrophilic dye, i.e., calcein. The plasma-induced release of calcein from liposomes is then used as a measure of liposome membrane integrity and, consequently, interaction between the cold atmospheric plasma and lipid bilayers. Physical mechanisms leading to membrane disruption are suggested, based on the plasma characterization including gas temperature calculation.

Nonthermal plasma technology for organic destruction

International Nuclear Information System (INIS)

Heath, W.O.; Birmingham, J.G.

1995-01-01

Pacific Northwest Laboratory (PNL) is investigating the use of nonthermal, electrically driven plasmas for destroying organic contaminants near ambient temperatures and pressures. Three different plasma systems have been developed to treat organics in air, water, and soil. These systems are the gas-phase corona reactor (GPCR) for treating air, the liquid phase corona reactor for treating water, and the in-situ corona for treating soils. This paper focuses on the GPCR as an alternative to other air purification technologies for treating off-gasses from remedial action efforts and industrial emissions

Tokamak plasma equilibrium problems with anisotropic pressure and rotation and their numerical solution

International Nuclear Information System (INIS)

Ivanov, A. A.; Martynov, A. A.; Medvedev, S. Yu.; Poshekhonov, Yu. Yu.

2015-01-01

In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (with arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented

Nonthermal plasma reactors for the production of light hydrocarbon olefins from heavy oil

Directory of Open Access Journals (Sweden)

G. Prieto

2003-03-01

Full Text Available During the last decade, nonthermal plasma technology was applied in many different fields, focusing attention on the destruction of harmful compounds in the air. This paper deals with nonthermal plasma reactors for the conversion of heavy oil into light hydrocarbon olefins, to be employed as gasoline components or to be added in small amounts for the catalytic reduction of nitrogen oxide compounds in the treatment of exhaust gas at power plants. For the process, the plate-plate nonthermal plasma reactor driven by AC high voltage was selected. The reactor was modeled as a function of parameter characteristics, using the methodology provided by the statistical experimental design. The parameters studied were gap distance between electrodes, carrier gas flow and applied power. Results indicate that the reactions occurring in the process of heavy oil conversion have an important selective behavior. The products obtained were C1-C4 hydrocarbons with ethylene as the main compound. Operating the parameters of the reactor within the established operative window of the system and close to the optimum conditions, efficiencies as high as 70 (mul/joule were obtained. These values validate the process as an in-situ method to produce light olefins for the treatment of nitrogen oxides in the exhaust gas from diesel engines.

Atmospheric-Pressure Non-Thermal Plasma Jet for biomedical and industrial applications

International Nuclear Information System (INIS)

Asenjo, J; Mora, J; Vargas, A; Brenes, L; Montiel, R; Arrieta, J; Vargas, VI

2015-01-01

In this work we present the development and evaluation of a low-cost DBD Plasma- JET reactor using Argon as carrier gas, this device is capable of generating a cold plasma plume several centimeters in length making it suitable for use directly in contact with objects and delicate materials, including living tissue. (paper)

Tritium issues in commercial pressurized water reactors

International Nuclear Information System (INIS)

Jones, G.

2008-01-01

Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

International Nuclear Information System (INIS)

Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

1988-01-01

This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)