Natural convection and wall radiation in tall cavities

International Nuclear Information System (INIS)

Balaji, C.; Venkateshan, S.P.

1996-01-01

The problem of combined natural convection and wall radiation in tall cavities has been taken up for a detailed numerical investigation. The governing equations for fluid flow have been solved by a finite volume method and the radiation has been treated by the radiosity-irradiation method. The analysis has been specifically made for the case where the emissivity of the hot left wall is different from that of the cold right wall. For this case it was found that decoupling radiation from free convection can lead to considerable error. Correlations have been suggested for predicting both the convective as well as the radiative heat transfer rates across the cavity. (author). 7 refs., 3 figs., 3 tabs

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.

Nuclear reactor neutron shielding

Science.gov (United States)

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

2017-09-12

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

Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

1981-01-01

Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

1981-01-01

Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

International Nuclear Information System (INIS)

Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

1981-01-01

Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

Nuclear reactor cavity floor passive heat removal system

Science.gov (United States)

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

2018-03-06

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

Gas dynamics in the central cavity of HYLIFE-II reactor

International Nuclear Information System (INIS)

Chen, X.M.; Schrock, V.E.; Peterson, P.F.; Colella, P.

1992-01-01

In a HYLIFE-II ICF reactor, the microfusion of the D-T capsule in the center of the chamber produces X-rays that can ablate a thin layer off the liquid blanket which protects the first structural wall Thisablated material will implode toward the center line of the central cavity due to the initial vacuum and cylindrical geometry, and then rebound back to the liquid blanket vent through it and exert a pressure ''impulse'' onto the structural wall. The initial ablation occurs in a very short period with very small characteristic length and the implosion and rebounding processes feature very high pressures and temperatures. The proper design of the chamber relies on the reasonably accurate analysis of the gas dynamics in the central cavity and the gas-liquid interaction. In this paper, a second order Godunov numerical method is used to solve the compressible flow equations in the central cavity. The rarefaction and shock phenomena are very well captured by the numerical calculation. The equation of state for Flibe vapor is used in the calculation along with the parameters for the HYLIFE-II design. Since the radiation transport has not yet been included in the current calculations, the vapor possesses higher energy and therefore temperature. The total mass vaporized will also be underestimated in the later time of the calculation. The incorporation of a radiation calculation into this code is our next goal

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

X-ray and pressure conditions on the first wall of a particle beam inertial confinement reactor

International Nuclear Information System (INIS)

Magelssen, G.R.

1979-01-01

Because of the presence of a chamber gas in a particle beam reactor cavity, nonneutron target debris created from thermonuclear burn will be modified or stopped before it reaches the first reactor wall. The resulting modified spectra and pulse lengths of the debris need to be calculated to determine first wall effects. Further, the cavity overpressure created by the momentum and energy exchange between the debris and gas must also be calculated to determine its effect. The purpose of this paper is to present results of the debris-background gas problem obtained with a one fluid, two temperature plasma hydrodynamic computer code model which includes multifrequency radiation transport. Spherical symmetry, ideal gas equation of state, and LTE for each radiation frequency group were assumed. The transport of debris ions was not included and all the debris energy was assumed to be in radiation. The calculated x-ray spectra and pulse lengths and the background overpressure are presented

A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

International Nuclear Information System (INIS)

Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

2015-01-01

A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

BEACON/MOD2A analysis of the Arkansas-1 reactor cavity during a hypothetical hot leg break

International Nuclear Information System (INIS)

Ramsthaler, J.A.

1979-01-01

As part of the evaluation of the new MOD2A version of the BEACON code, the Arkansas-1 reactor cavity was modeled during a hypothetical loss-of-coolant accident. Results of the BEACON analysis were compared with results obtained previously with the COMPARE containment code. Studies were also made investigating some of the BEACON interphasic, timestep control, and wall heat transfer options to assure that these models were working properly and to observe their effects on the results. Descriptions of the Arkansas-1 reactor cavity, initial assumptions during the hypothetical LOCA, and methods of modeling with BEACON are presented. Some of the problems encountered in accurately modeling the penetrations surrounding the hot and cold leg pipes are also discussed

Chamber wall response to target implosion in inertial fusion reactors : new and critical assessments

International Nuclear Information System (INIS)

Hassanein, A.; Morozov, V.

2002-01-01

The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The hydrodynamic response of gas filled cavities and photon radiation transport of the deposited energy has been calculated by means of new and advanced numerical techniques. Fragmentation models of liquid jets as a result of the deposited energy have also been developed, and the impact on chamber clearing dynamics has been evaluated. Th focus of this study is to critically assess the reliability and the dynamic response of chamber walls in various proposed protection methods for IFE systems. Of particular concern is the effect on wall erosion lifetime of various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls

Feasibility of flooding the reactor cavity with liquid gallium coolant for IVR-ERVC strategy

International Nuclear Information System (INIS)

Park, Seong Dae; Bang, In Cheol

2013-01-01

Highlights: ► We investigate the feasibility of gallium liquid metal application for IVR-ERVC. ► We consider overall concerns to apply the liquid metal. ► Decay heat can be removed by flooding the reactor cavity with gallium liquid metal. -- Abstract: In this paper, a new approach replacing the ERVC coolant by a liquid metal instead of water is studied to avoid the heat removal limit of CHF during boiling of water. As the flooding material, gallium is used in terms of the melting and boiling points. Gallium has the enough low melting point of ∼29.7 °C to ensure to maintain liquid state within the containment building. A gallium storage tank for the new flooding system of the ERVC is located in higher position than one of the reactor cavity to make a passive system using the gravity for the event of a station blackout (SBO). While the decay heat from the reactor vessel is removed by gallium, the borated water which is coming out from the reactor system plays a role as the ultimate heat sink in this ERVC system. In the system, two configurations of gallium and borated water are devised depending on whether the direct contact between them occurs. In the first configuration, two fluids are separated by the block structure. The decay heat is transported from molten corium to gallium through the vessel wall. Then the heat is ultimately dissipated by boiling of water in the block structure surface facing the borated water. In the second configuration, the cavity is flooded with both borated water and gallium in the same reactor cavity space. As the result, two layers of the fluids are naturally formed by the density difference. Like the first configuration, finally the heat removal is achieved by boiling of water via gallium. The CFD analysis shows that the maximum temperature of gallium is much lower than its boiling point while the natural circulation is stably formed in two types of the configurations without any serious risk of thermal limit

Simulation of IVR-ERVC and estimation method of coolant inflow to the cavity

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyunjin; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

In this study, the temperature distribution outside of RV wall and evaporation rate due to heat from core will be investigated. Using the universal analysis program ANSYS Fluent, the natural convection in the cavity for IVR-ERVC conditions were modelled and performed for heat transfer analysis. The aim of this study is to calculate the appropriate coolant flow so that coolant level in the cavity can be maintained at prescribed level and vessel wall temperature distribution, including RV outside wall temperature are also investigated. Reactor vessel and cavity in case of ex-vessel cooling for severe accident condition were modeled with and without insulators. The heat load into reactor vessel from corium inside of reactor lower head were obtained from MELCORE analysis and used as input B.C of CFD analysis. The Temperature gradient of reactor outer surface and evaporation rate of cooling eater was obtained from the analysis. These results can be used for further analysis of reactor vessel creep behavior and the estimate the coolant flow rate into the reactor cavity.. and The result can be used to verify the natural convection phenomena in the cavity and also to set the design parameters of cavity and coolant flow rate. The vessel outer surface temperature gradient can be also used to more accurate investigation of vessel creep behavior during severe accident condition, The result can also be used set up a strategy for severe accident managements.

Permanent cavity seal ring for a nuclear reactor containment arrangement

International Nuclear Information System (INIS)

Swidwa, K.J.; Salton, R.B.; Marshall, J.R.

1990-01-01

This patent describes a nuclear reactor containment arrangement. It comprises: a reactor pressure vessel which thermally expands and contracts during cyclic operation of the reactor, the vessel having a peripheral wall and a horizontally outwardly extending flange thereon; a containment wall having a shelf, the wall spaced from and surrounding the peripheral wall of the reactor pressure vessel defining an annular expansion gap therebetween, and an annular ring seal extending across the annular expansion gap to provide a water-tight seal therebetween

Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

Science.gov (United States)

Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

2018-03-01

In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

Integrity of the first wall in fusion reactors

International Nuclear Information System (INIS)

Kurihara, Ryoichi

2004-07-01

Future fusion power reactors DREAM and A-SSTR2, which have been conceptually designed in the Japan Atomic Energy Research Institute, use the SiC/SiC composite material as the first wall of the blanket because of its characteristics of high heat-resistance and low radiation material. DEMO reactor, which was conceptually designed in 2001, uses the low activation ferritic steel as the first-wall material of the blanket. The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel. (author)

1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

International Nuclear Information System (INIS)

Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

1994-11-01

The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in x 1.2 m x 17.1 cm thick [4 ft x 4 ft x 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the open-quotes mirrorclose quotes insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in x 2.1 in [10 ft x 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28 degrees C/hr [12.5, 25, and 50 degrees F/hr] as measured on the heated face. A peak temperature of 454 degrees C [850 degrees F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing

A two-cavity reactor for solar chemical processes: heat transfer model and application to carbothermic reduction of ZnO

International Nuclear Information System (INIS)

Wieckert, Christian; Palumbo, Robert; Frommherz, Ulrich

2004-01-01

A 5 kW two-cavity beam down reactor for the solar thermal decomposition of ZnO with solid carbon has been developed and tested in a solar furnace. Initial exploratory experiments show that it operates with a solar to chemical energy conversion efficiency of about 15% when the solar flux entering the reactor is 1300 kW/m 2 , resulting in a reaction chamber temperature of about 1500 K. The solid products have a purity of nearly 100% Zn. Furthermore, the reactor has been described by a numerical model that combines radiant and conduction heat transfer with the decomposition kinetics of the ZnO-carbon reaction. The model is based on the radiosity exchange method. For a given solar input, the model estimates cavity temperatures, Zn production rates, and the solar to chemical energy conversion efficiency. The model currently makes use of two parameters which are determined from the experimental results: conduction heat transfer through the reactor walls enters the model as a lumped term that reflects the conduction loss during the experiments, and the rate of the chemical reaction includes an experimentally determined term that reflects the effective amount of ZnO and CO participating in the reactor. The model output matches well the experimentally determined cavity temperatures. It suggests that reactors built with this two-cavity concept already on this small scale can reach efficiencies exceeding 25%, if operated with a higher solar flux or if one can reduce conduction heat losses through better insulation and if one can maintain or improve the effective amount of ZnO and CO that participates in the reaction

Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

Chamber wall response to target implosion in inertial fusion reactors: new and critical assessments

International Nuclear Information System (INIS)

Hassanein, A.; Morozov, V.

2002-01-01

The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to next target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The focus of this study is to critically assess the reliability and the dynamic response of chamber walls in IFE systems. Of particular concern is the effect on wall erosion lifetime due to various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls

Wall compliance and violin cavity modes.

Science.gov (United States)

Bissinger, George

2003-03-01

Violin corpus wall compliance, which has a substantial effect on cavity mode frequencies, was added to Shaw's two-degree-of-freedom (2DOF) network model for A0 ("main air") and A1 (lowest length mode included in "main wood") cavity modes. The 2DOF model predicts a V(-0.25) volume dependence for A0 for rigid violin-shaped cavities, to which a semiempirical compliance correction term, V(-x(c)) (optimization parameter x(c)) consistent with cavity acoustical compliance and violin-based scaling was added. Optimizing x(c) over A0 and A1 frequencies measured for a Hutchins-Schelleng violin octet yielded x(c) approximately 0.08. This markedly improved A0 and A1 frequency predictions to within approximately +/- 10% of experiment over a range of about 4.5:1 in length, 10:1 in f-hole area, 3:1 in top plate thickness, and 128:1 in volume. Compliance is a plausible explanation for A1 falling close to the "main wood" resonance, not increasingly higher for the larger instruments, which were scaled successively shorter compared to the violin for ergonomic and practical reasons. Similarly incorporating compliance for A2 and A4 (lowest lower-/upper-bout modes, respectively) improves frequency predictions within +/-20% over the octet.

Frequency-tunable SRF cavities for microwave opto-mechanics

Science.gov (United States)

Castelli, Alessandro; Martinez, Luis; Pate, Jacob; Thompson, Johnathon; Chiao, Raymond; Sharping, Jay

Three dimensional SRF (Superconducting Radio Frequency) cavities are known for achieving high quality factors (Q =109 or higher) but suffer from limited frequency tunability once fabricated and cooled to superconducting temperatures. Our end-wall design allows for numerous applications of cavity tuning at temperatures as low as 40 millikelvin. Using a bimorphic piezoelectric transducer, we demonstrate approximately 15 MHz of resonance tunability for the TE011 mode at cryogenic temperatures in a cylindrical reactor grade niobium (Nb) cavity (10% of the range at room temperature). This range doubles when using tunable end-walls on both cavity ends. We report on techniques for improving the Q of multi-component cavities including the use of concave end-walls to reduce fields near the cylinder ends and indium O-rings to reduce resistive losses at the gaps. Three-dimensional SRF cavities of this type have potential applications to quantum information science, precision displacement metrology, and quantum electro-dynamics.

Dismantling method for reactor shielding wall and device therefor

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1995-01-01

A ring member having an outer diameter slightly smaller than an inner diameter of a reactor shielding wall to be dismantled is lowered in the inside of the reactor shielding wall while keeping a horizontal posture. A cutting device is disposed at the lower peripheral edge of the ring member. The cutting device can move along the peripheral edge of the circular shape of the ring member. The ring member is urged against the inner surface of the reactor shielding wall by using an urging member to immobilize the ring member. Then, the cutting device is operated to cut the reactor shielding wall into a plurality of ring-like blocks at a plurality of inner horizontal ribs or block connection ribs. Then, the blocks of the cut reactor shielding wall are supported by the ring member, and transported out of the reactor container by a lift. The cut blocks transported to the outside are finely dismantled for every block in a closed chamber. (I.N.)

Reactor wall in thermonuclear device

International Nuclear Information System (INIS)

Shibui, Masanao.

1988-01-01

Purpose: To always monitor the life of armours in reactor walls and automatically shutdown the reactor if it should be operated in excess of the limit of use. Constitution: Monitoring material of lower melting point than armours (for example beryllium pellets) as one of the reactor wall constituents of a thermonuclear device are embedded in a region leaving the thickness corresponding to the allowable abrasion of the armour. In this structure, if the armours are abrased due to particle loads of a plasma and the abrasion exceeds a predetermined allowable level, the monitoring material is exposed to the plasma and melted and evaporated. Since this can be detected by impurity monitors disposed in the reactor, it is possible to recognize the limit for the working life of the armours. If the thermonuclear reactor should be operated accidentally exceeding the life of the armours, since a great amount of the monitoring materials have been evaporated, they flow into the plasma to increase the plasma radiation loss thereby automatically eliminate the plasma. (K.M.)

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

International Nuclear Information System (INIS)

Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

2014-01-01

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

Energy Technology Data Exchange (ETDEWEB)

Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

2014-07-14

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

Apparatus for treating the walls and floor of the pelvic cavity with radiation

International Nuclear Information System (INIS)

Clayton, R.S.

1975-01-01

An apparatus for reaing carcinoma of the walls and floor of the pelvic cavity is described. An elongated tube has an inner end adapted to be placed in the pelvic cavity and an outer end adapted to extend through to the outside of the body. Radioactive material is placed at the inner end. An inner balloon above the radioactive material is inflated to hold a body of liquid shielding material such as mercury. A lower balloon portion beneath the inner balloon spaces areas to be treated such as the walls and floor of the pelvic cavity from the radioactive material. An upper balloon portion above the inner balloon keeps the intestines out of the pelvic cavity and away from the radioactive material. The apparatus is inserted into the pelvic cavity through an abdominal incision. When treating a woman for carcinoma in the walls and floor of the pelvic cavity the tube is moved through the vaginal passage from the inside outwardly. When treating a woman with a closed vaginal passage, as may result from surgery, or when treating a man, such as for carcinoma of the bladder, the tube will pass out of the body through a lower abdominal incision. Following treatment, all balloons are deflated so that the apparatus can be withdrawn through the vaginal passage or the lower abdominal incision, as the case may be. (auth)

Computer-controlled wall servicing robot

Energy Technology Data Exchange (ETDEWEB)

Lefkowitz, S. [Pentek, Inc., Corapolis, PA (United States)

1995-03-01

After four years of cooperative research, Pentek has unveiled a new robot with the capability to automatically deliver a variety of cleaning, painting, inspection, and surveillance devices to large vertical surfaces. The completely computer-controlled robot can position a working tool on a 50-foot tall by 50-foot wide vertical surface with a repeatability of 1/16 inch. The working end can literally {open_quotes}fly{close_quotes} across the face of a wall at speed of 60 per minute, and can handle working loads of 350 pounds. The robot was originally developed to decontaminate the walls of reactor fueling cavities at commercial nuclear power plants during fuel outages. If these cavities are left to dry after reactor refueling, contamination present in the residue could later become airborne and move throughout the containment building. Decontaminating the cavity during the refueling outage reduces the need for restrictive personal protective equipment during plant operations to limit the dose rates.

Heat transfer in reactor cavity during core-concrete interaction

International Nuclear Information System (INIS)

Adroguer, B.; Cenerino, G.

1989-08-01

In the unlikely event of a severe accident in a nuclear power plant, the core may melt through the vessel and slump into the concrete reactor cavity. The hot mixture of the core material called corium interacts thermally with the concrete basemat. The WECHSL code, developed at K.f.K. Karlsruhe in Germany is used at the Protection and Nuclear Safety Institute (I.P.S.N.) of CEA to compute this molten corium concrete interaction (MCCI). Some uncertainties remain in the partition of heat from the corium between the basemat and the upper surrounding structures in the cavity where the thermal conditions are not computer. The CALTHER code, under development to perform a more mechanistic evaluation of the upward heat flux has been linked to WECHSL-MOD2 code. This new version enables the modelling of the feedback effects from the conditions in the cavity to the MCCI and the computation of the fraction of upward flux directly added to the cavity atmosphere. The present status is given in the paper. Preliminary calculations of the reactor case for silicate and limestone common sand (L.C.S.) concretes are presented. Significant effects are found on concrete erosion, gases release and temperature of the upper part of corium, particularly for L.C.S. concrete

Reactor scale modeling of multi-walled carbon nanotube growth

International Nuclear Information System (INIS)

Lombardo, Jeffrey J.; Chiu, Wilson K.S.

2011-01-01

As the mechanisms of carbon nanotube (CNT) growth becomes known, it becomes important to understand how to implement this knowledge into reactor scale models to optimize CNT growth. In past work, we have reported fundamental mechanisms and competing deposition regimes that dictate single wall carbon nanotube growth. In this study, we will further explore the growth of carbon nanotubes with multiple walls. A tube flow chemical vapor deposition reactor is simulated using the commercial software package COMSOL, and considered the growth of single- and multi-walled carbon nanotubes. It was found that the limiting reaction processes for multi-walled carbon nanotubes change at different temperatures than the single walled carbon nanotubes and it was shown that the reactions directly governing CNT growth are a limiting process over certain parameters. This work shows that the optimum conditions for CNT growth are dependent on temperature, chemical concentration, and the number of nanotube walls. Optimal reactor conditions have been identified as defined by (1) a critical inlet methane concentration that results in hydrogen abstraction limited versus hydrocarbon adsorption limited reaction kinetic regime, and (2) activation energy of reaction for a given reactor temperature and inlet methane concentration. Successful optimization of a CNT growth processes requires taking all of those variables into account.

Gas-core reactor power transient analysis. Final report

International Nuclear Information System (INIS)

Kascak, A.F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

Segmented trapped vortex cavity

Science.gov (United States)

Grammel, Jr., Leonard Paul (Inventor); Pennekamp, David Lance (Inventor); Winslow, Jr., Ralph Henry (Inventor)

2010-01-01

An annular trapped vortex cavity assembly segment comprising includes a cavity forward wall, a cavity aft wall, and a cavity radially outer wall there between defining a cavity segment therein. A cavity opening extends between the forward and aft walls at a radially inner end of the assembly segment. Radially spaced apart pluralities of air injection first and second holes extend through the forward and aft walls respectively. The segment may include first and second expansion joint features at distal first and second ends respectively of the segment. The segment may include a forward subcomponent including the cavity forward wall attached to an aft subcomponent including the cavity aft wall. The forward and aft subcomponents include forward and aft portions of the cavity radially outer wall respectively. A ring of the segments may be circumferentially disposed about an axis to form an annular segmented vortex cavity assembly.

Albedo analytical method for multi-scattered neutron flux calculation in cavity

International Nuclear Information System (INIS)

Shin, Kazuo; Selvi, S.; Hyodo, Tomonori

1986-01-01

A simple formula which describes multi-scattered neutron flux in a spherical cavity was derived based on the albedo concept. The formura treats a neutron source which has an arbitrary energy-angle distribution and is placed at any point in the cavity. The derived formula was applied to the estimation of neutron fluxes in two cavities, i.e. a spherical concrete cell with a 14-MeV neutron source at the center and the ''YAYOI'' reactor cavity with a pencil beam of reactor neutrons. The results of the analytical formula agreed very well with the reference data in the both problems. It was concluded that the formula is applicable to estimate the neutron fluxes in a spherical cell except for special cases that tangential source neutrons are incident to the cavity wall. (author)

Comparison of BEACON and COMPARE reactor cavity subcompartment analyses

International Nuclear Information System (INIS)

Burkett, M.W.; Idar, E.S.; Gido, R.G.; Lime, J.F.; Koestel, A.

1984-04-01

In this study, a more advanced best-estimate containment code, BEACON-MOD3A, was ued to calculate force and moment loads resulting from a high-energy blowdown for two reactor cavity geometries previously analyzed with the licensing computer code COMPARE-MOD1A. The BEACON force and moment loads were compared with the COMPARE results to determine the safety margins provided by the COMPARE code. The forces and moments calculated by the codes were found to be different, although not in any consistent manner, for the two reactor cavity geometries studied. Therefore, generic summary statements regarding margins cannot be made because of the effects of the detailed physical configuration. However, differences in the BEACON and COMPARE calculated forces and moments can be attributed to differences in the modeling assumptions used in the codes and the analyses

Magnesium diboride on inner wall of copper tube: A test case for superconducting radio frequency cavities

Science.gov (United States)

Withanage, Wenura K.; Lee, N. H.; Penmatsa, Sashank V.; Wolak, M. A.; Nassiri, A.; Xi, X. X.

2017-10-01

Superconductor magnesium diboride is considered one of the viable materials to substitute bulk niobium for superconducting radio frequency cavities. Utilizing a MgB2 coating on the inner wall of a copper cavity will allow operation at higher temperatures (20-25 K) than Nb cavities due to the high transition temperature of MgB2 (39 K) and the high thermal conductivity of Cu. In this paper, we present results of MgB2 coating on Cu tubes with similar dimensions to a 3 GHz cavity, as the first step towards coating the actual cavity, using the hybrid physical chemical vapor deposition technique. The results show successful coating of a uniform MgB2 layer on the inner wall of the Cu tubes with Tc as high as 37 K.

Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

1997-01-01

For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

Considerations relating to the presence of water in the reactor cavity during severe accidents

International Nuclear Information System (INIS)

Perez, F.; Morales, M.D.

1994-01-01

The purpose of this paper is to present some of the factors, both positive and negative, associated with the presence of water in the reactor cavity. The presence of water in the reactor cavity is one of the factors whose influence on the evolution of severe accidents must be determined since, on the one hand, it has an impact on some of the most significant severe accident phenomena and, on the other, it could be an important factor when preparing accident management strategies resulting from containment analyses. In spite of the initial intuitive impression that water in the reactor cavity must always be beneficial, certain phenomena, such as the following must also be taken into account before developing accident management strategies: - Higher production of steam - Possibility of steam explosions - Increased production of H 2 due to oxidation of steel components of the melted core ejected from the vessel - More oxidation energy released due to the presence of oxygen in the cavity (Author)

Magnesium diboride on inner wall of copper tube: A test case for superconducting radio frequency cavities

Directory of Open Access Journals (Sweden)

Wenura K. Withanage

2017-10-01

Full Text Available Superconductor magnesium diboride is considered one of the viable materials to substitute bulk niobium for superconducting radio frequency cavities. Utilizing a MgB_{2} coating on the inner wall of a copper cavity will allow operation at higher temperatures (20–25 K than Nb cavities due to the high transition temperature of MgB_{2} (39 K and the high thermal conductivity of Cu. In this paper, we present results of MgB_{2} coating on Cu tubes with similar dimensions to a 3 GHz cavity, as the first step towards coating the actual cavity, using the hybrid physical chemical vapor deposition technique. The results show successful coating of a uniform MgB_{2} layer on the inner wall of the Cu tubes with T_{c} as high as 37 K.

Neutron and gamma characterization within the FFTF reactor cavity

International Nuclear Information System (INIS)

Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

1980-08-01

Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

International Nuclear Information System (INIS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

Material options for a commercial fusion reactor first wall

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m 2 . A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW 2 , provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys

Heat transfer in inertial confinement fusion reactor systems

International Nuclear Information System (INIS)

Hovingh, J.

1979-01-01

The transfer of energy produced by the interaction of the intense pulses of short-ranged fusion microexplosion products with materials is one of the most difficult problems in inertially-confined fusion (ICF) reactor design. The short time and deposition distance for the energy results in local peak power densities on the order of 10 18 watts/m 3 . High local power densities may cause change of state or spall in the reactor materials. This will limit the structure lifetimes for ICF reactors of economic physical sizes, increasing operating costs including structure replacement and radioactive waste management. Four basic first wall protection methods have evolved: a dry-wall, a wet-wall, a magnetically shielded wall, and a fluid wall. These approaches are distinguished by the way the reactor wall interfaces with fusion debris as well as the way the ambient cavity conditions modify the fusion energy forms and spectra at the first wall. Each of these approaches requires different heat transfer considerations

Analysis of reactor cavity radiation streaming: some practical considerations

International Nuclear Information System (INIS)

Simmons, G.L.

1979-01-01

A description is presented of a cost effective analysis procedure for use in the prediction of radiation environments in the cavity and containment building of a nuclear power reactor. Comments are offered on potential problems in certification of analysis procedures and the availability of benchmarkable data sets, both measurements and calculations

Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

International Nuclear Information System (INIS)

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

1985-04-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

First wall studies of a laser-fusion hybrid reactor design

International Nuclear Information System (INIS)

Hovingh, J.

1976-09-01

The design of a first wall for a 20 MW thermonuclear power laser fusion hybrid reactor is presented. The 20 mm thick graphite first wall is located 3.5 m from the DT microexplosion with a thermonuclear yield of 10 MJ. Estimates of the energy deposition, temperature, stresses, and material vaporized from the first wall due to the interaction of the x-rays, charged particle debris, and reflected laser light with the graphite are presented, along with a brief description of the analytical methods used for these estimations. Graphite is a viable first wall material for inertially-confined fusion reactors, with lifetimes of a year possible

Heavy ion beam transport through liquid lithium first wall ICF reactor cavities

International Nuclear Information System (INIS)

Stroud, P.D.

1985-01-01

This analysis addresses the critical issue of the final transport of a heavy ion beam in an inertial confinement fusion reactor. The beam must traverse the reaction chamber from the final focusing lens to the target without being disrupted. This requirement has a strong impact on the reactor design. It is essential to the development of ICF fusion reactor technology, that the restrictions placed on the reactor engineering parameters by final beam transport consideration be understood early on

Electromagnetic effects involving a tokamak reactor first wall and blanket

International Nuclear Information System (INIS)

Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

1980-01-01

Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

Scale-up of microwave assisted flow synthesis by transient processing through monomode cavities in series

NARCIS (Netherlands)

Patil, N.G.; Benaskar, F.; Rebrov, E.; Meuldijk, J.; Hulshof, L.A.; Hessel, V.; Schouten, J.C.

2014-01-01

A new scale-up concept for microwave assisted flow processing is presented where modular scale-up is achieved by implementing microwave cavities in series. The scale-up concept is demonstrated for case studies of a packed-bed reactor and a wall-coated tubular reactor. With known kinetics and

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

1986-01-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

Continuously renewed wall for a thermonuclear reactor

International Nuclear Information System (INIS)

Livshits, A.I.; Pustovojt, YU.M.; Samartsev, A.A.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii)

1982-01-01

The possibility of creating a continuously renewed first wall of a thermonuclear reactor is experimentally investigated. The following variants of the wall are considered: the wall is double, its part turned to plasma is made of comparatively thin material. The external part separated from it by a small gap appears to be protected from interaction with plasma and performs structural functions. The gap contains the mixture of light helium and hydrogen and carbon-containing gas. The light gas transfers heat from internal part of the wall to the external part. Carbon-containing gas provides continuous renewal of carbon coating of the operating surface. The experiment is performed with palladium membrane 20 μm thick. Carbon is introduced into the membrane by benzol pyrolysis on one of the surfaces at the membrane temperature of 900 K. Carbon removal from the operating side of the wall due to its spraying by fast particles is modelled by chemical itching with oxygen given to the operating membrane wall. Observation of the carbon release on the operating surface is performed mass-spectrometrically according to the observation over O 2 transformation into CO and CO 2 . It is shown that in cases of benzol pressure of 5x10 -7 torr, carbon current on the opposite surface is not less than 3x10 12 atoms/sm 2 s and corresponds to the expected wall spraying rate in CF thermonuclear reactors. It is also shown that under definite conditions the formation and maintaining of a through protective carbon coating in the form of a monolayer or volumetric phase is possible

Structure of thermonuclear reactor wall

International Nuclear Information System (INIS)

Yamazaki, Seiichiro.

1991-01-01

In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

Tutu, N.K.; Ginsberg, T.; Klages, J.R.

1991-01-01

This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

1985-01-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

Casimir effect for closed cavities with conducting and permeable walls

International Nuclear Information System (INIS)

Ferreira, L.A.; Zimerman, A.H.; Ruggiero, J.R.

1980-01-01

The quantum electromagnetic zero point energy is calculated for rectangular cavities where some of the walls are perfect conductors and the others are made of infinitely permeable materials. It is found that for cubic systems, for some configurations the zero point electromagnetic energy is positive, while in other configurations this zero point energy is negative. The consequences of these results on possible models for the electron are discussed. (Author) [pt

Heat transfer of natural convection in a rectangular cavity with vertical walls of different temperatures

International Nuclear Information System (INIS)

Seki, Nobuhiro; Fukusako, Shoichiro; Inaba, Hideo

1978-01-01

In the present study the behavior of heat transfer in a rectangular cavity with one isothermal vertical wall heated and the other cooled is investigated. Heat transfer coefficients on the vertical walls are measured for fluids with Prandtl number Pr of 3 to 40,000 in case of aspect-ratio H/W from 5 to 47.5 and their correlated results are presented for laminar, transition and turbulent regions, respectively. It is shown that the present arrangement (Nu sub(H) - Ra sub(H)) using the height of cavity as a representative length may significantly be useful in the various heat transfer modes accompanied with flow patterns of them. (auth.)

Thermal resistances of air in cavity walls and their effect upon the thermal insulation performance

Energy Technology Data Exchange (ETDEWEB)

Bekkouche, S.M.A.; Cherier, M.K.; Hamdani, M.; Benamrane, N. [Application of Renewable Energies in Arid and Semi Arid Environments /Applied Research Unit on Renewable Energies/ EPST Development Center of Renewable Energies, URAER and B.P. 88, ZI, Gart Taam Ghardaia (Algeria); Benouaz, T. [University of Tlemcen, BP. 119, Tlemcen R.p. 13000 (Algeria); Yaiche, M.R. [Development Center of Renewable Energies, CDER and B.P 62, 16340, Route de l' Observatoire, Bouzareah, Algiers (Algeria)

2013-07-01

The optimum thickness in cavity walls in buildings is determined under steady conditions; the heat transfer has been calculated according to ISO 15099:2003. Two forms of masonry units are investigated to conclude the advantage of high thermal emissivity. The paper presents also some results from a study of the thermal insulation performance of air cavities bounded by thin reflective material layer 'eta = 0.05'. The results show that the most economical cavity configuration depends on the thermal emissivity and the insulation material used.

Effect of finite cavity width on flow oscillation in a low-Mach-number cavity flow

Energy Technology Data Exchange (ETDEWEB)

Zhang, Ke; Naguib, Ahmed M. [Michigan State University, East Lansing, MI (United States)

2011-11-15

The current study is focused on examining the effect of the cavity width and side walls on the self-sustained oscillation in a low Mach number cavity flow with a turbulent boundary layer at separation. An axisymmetric cavity geometry is employed in order to provide a reference condition that is free from any side-wall influence, which is not possible to obtain with a rectangular cavity. The cavity could then be partially filled to form finite-width geometry. The unsteady surface pressure is measured using microphone arrays that are deployed on the cavity floor along the streamwise direction and on the downstream wall along the azimuthal direction. In addition, velocity measurements using two-component Laser Doppler Anemometer are performed simultaneously with the array measurements in different azimuthal planes. The compiled data sets are used to investigate the evolution of the coherent structures generating the pressure oscillation in the cavity using linear stochastic estimation of the velocity field based on the wall-pressure signature on the cavity end wall. The results lead to the discovery of pronounced harmonic pressure oscillations near the cavity's side walls. These oscillations, which are absent in the axisymmetric cavity, are linked to the establishment of a secondary mean streamwise circulating flow pattern near the side walls and the interaction of this secondary flow with the shear layer above the cavity. (orig.)

Nuclear reactor vessel fuel thermal insulating barrier

Science.gov (United States)

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Reactor pressure vessel failure probability following through-wall cracks due to pressurized thermal shock events

International Nuclear Information System (INIS)

Simonen, F.A.; Garnich, M.R.; Simonen, E.P.; Bian, S.H.; Nomura, K.K.; Anderson, W.E.; Pedersen, L.T.

1986-04-01

A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events

RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Chang Oh; Cliff Davis; Goon C. Park

2007-01-01

A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

Analysis of eddy currents in the walls of the ferrite tuned RF cavity for the TRIUMF Kaon factory booster synchrotron

International Nuclear Information System (INIS)

Enchevich, I.B.; Barnes, M.J.; Poirier, R.L.

1991-05-01

In the perpendicular biased ferrite tuned cavity of the proposed TRIUMF Kaon Factory Booster Synchrotron, magnetizing flux passes through the cavity walls. If special care is not taken to minimize eddy current loss in the walls, the dissipated power would be excessive and the magnetic fields set up by the eddy currents would disturb the magnetic field being applied. By electrically isolating the cooling structure from the cavity walls and introducing slots in the walls it is possible to bring to an acceptable level both the power loss and the maximal temperatures. Based on the measurements, an analytical model - essentially 3D - was derived and the eddy currents were predicted using the circuit analysis program PSpice. The calculated surface current and power distribution agree with measurements. PSpice can now be used to determine the effect of design changes on the eddy current and power distribution. (Author) 7 refs., 5 figs

Reactor cavity streaming: the problem and engineered solutions

International Nuclear Information System (INIS)

Iotti, R.C.; Yang, T.L.; Rogers, W.H.

1979-01-01

Experience at operating pressurized water reactors has revealed that air gaps between the reactor vessel and the biological shield wall can provide paths for radiation streaming, which may prohibitively limit the accessibility required to areas in the containment during power operation, increase personnel exposure during shutdown, and cause radiation damage to equipment and cables located above the vessel. Several concepts of shield are discussed together with their predicted effectiveness. The analytical methods employed to determine the streaming magnitude and the shield effectiveness are also discussed and their accuracy is measured by comparison with actual measurement at an operating plant

Models and analyses for inertial-confinement fusion-reactor studies

International Nuclear Information System (INIS)

Bohachevsky, I.O.

1981-05-01

This report describes models and analyses devised at Los Alamos National Laboratory to determine the technical characteristics of different inertial confinement fusion (ICF) reactor elements required for component integration into a functional unit. We emphasize the generic properties of the different elements rather than specific designs. The topics discussed are general ICF reactor design considerations; reactor cavity phenomena, including the restoration of interpulse ambient conditions; first-wall temperature increases and material losses; reactor neutronics and hydrodynamic blanket response to neutron energy deposition; and analyses of loads and stresses in the reactor vessel walls, including remarks about the generation and propagation of very short wavelength stress waves. A discussion of analytic approaches useful in integrations and optimizations of ICF reactor systems concludes the report

Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

International Nuclear Information System (INIS)

Monsler, M.J.; Meier, W.R.

1981-02-01

The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate

Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Monsler, M.J.; Meier, W.R.

1981-02-01

The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate.

Conjugate heat transfer in a porous cavity filled with nano-fluids and heated by a triangular thick wall

International Nuclear Information System (INIS)

Chamkha, Ali J.; Ismael, Muneer A.

2013-01-01

The conjugate natural convection-conduction heat transfer in a square domain composed of nano-fluids filled porous cavity heated by a triangular solid wall is studied under steady-state conditions. The vertical and horizontal walls of the triangular solid wall are kept isothermal and at the same hot temperature Th. The other boundaries surrounding the porous cavity are kept adiabatic except the right vertical wall where it is kept isothermally at the lower temperature T c . Equations governing the heat transfer in the triangular wall and heat and nano-fluid flow, based on the Darcy model, in the nano-fluid-saturated porous medium together with the derived relation of the interface temperature are solved numerically using the over-successive relaxation finite-difference method. A temperature independent nano-fluids properties model is adopted. Three nano-particle types dispersed in one base fluid (water) are investigated. The investigated parameters are the nano-particles volume fraction φ (0-0.2), Rayleigh number Ra (10-1000), solid wall to base-fluid saturated porous medium thermal conductivity ratio K ro (0.44, 1, 23.8), and the triangular wall thickness D (0.1-1). The results are presented in the conventional form; contours of streamlines and isotherms and the local and average Nusselt numbers. At a very low Rayleigh number Ra = 10, a significant enhancement in heat transfer within the porous cavity with φ is observed. Otherwise, the heat transfer may be enhanced or deteriorated with φ depending on the wall thickness D and the Rayleigh number Ra. At high Rayleigh numbers and low conductivity ratios, critical values of D, regardless of 4, are observed and accounted. (authors)

Lithium adsorption by the first wall of fusion reactor-tokamak

International Nuclear Information System (INIS)

Bakunin, O.G.

1989-01-01

Lithium adsorption by the first wall of fusion reactor under stationary conditions and in the absence of chemical reactions is considered. Possibility of achieving 70% coating of the wall with lithium which can lead to sufficient decrease of sputtering is shown. 5 refs.; 5 figs

Thermal analysis of bulk filled composite resin polymerization using various light curing modes according to the curing depth and approximation to the cavity wall

Directory of Open Access Journals (Sweden)

Hoon-Sang Chang

2013-07-01

Full Text Available OBJECTIVE: The purpose of this study was to investigate the polymerization temperature of a bulk filled composite resin light-activated with various light curing modes using infrared thermography according to the curing depth and approximation to the cavity wall. MATERIAL AND METHODS: Composite resin (AeliteFlo, Bisco, Schaumburg, IL, USA was inserted into a Class II cavity prepared in the Teflon blocks and was cured with a LED light curing unit (Dr's Light, GoodDoctors Co., Seoul, Korea using various light curing modes for 20 s. Polymerization temperature was measured with an infrared thermographic camera (Thermovision 900 SW/TE, Agema Infra-red Systems AB, Danderyd, Sweden for 40 s at measurement spots adjacent to the cavity wall and in the middle of the cavity from the surface to a 4 mm depth. Data were analyzed according to the light curing modes with one-way ANOVA, and according to curing depth and approximation to the cavity wall with two-way ANOVA. RESULTS: The peak polymerization temperature of the composite resin was not affected by the light curing modes. According to the curing depth, the peak polymerization temperature at the depth of 1 mm to 3 mm was significantly higher than that at the depth of 4 mm, and on the surface. The peak polymerization temperature of the spots in the middle of the cavity was higher than that measured in spots adjacent to the cavity wall. CONCLUSION: In the photopolymerization of the composite resin, the temperature was higher in the middle of the cavity compared to the outer surface or at the internal walls of the prepared cavity.

First wall costs of an ion-beam fusion reactor

International Nuclear Information System (INIS)

Hovingh, J.

1977-08-01

This paper parametrically investigates the effects of microexplosion energy on the first wall costs of a 4000 MW/sub t/ ion-beam initiated, inertially confined fusion reactor for several first wall materials. The thermodynamic models and the results for microexplosion energies between 400 and 4000 MJ are presented. A solid stainless steel or a composite isotropic graphite over stainless steel first wall can operate for a year at a cost of 0.6 mills per kWh gross electric power output

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

Pailla, Henri; Schaller, Karl; Vidard, Michel.

1974-01-01

A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

Energy Technology Data Exchange (ETDEWEB)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

2001-03-15

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

International Nuclear Information System (INIS)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin; Park, Jae Hong

2001-03-01

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report

The mechanical performance of the fusion reactor first wall. Pt. 2

International Nuclear Information System (INIS)

Daenner, W.; Raeder, J.

1977-03-01

While the first part of this report was concerned with the steady-state mechanical analysis of the fusion reactor first wall, this part deals with the analysis based upon pulsed load conditions. In a first section we elaborate various solutions of the non-stationary heat conduction problem in plane geometry capable of describing the temperature response of the wall due to characteristic plasma pulse sequences. these solutions are input to a quasi-steady-state stress and strain analysis. Finally, the results of this analysis are set in relation to the fatigue properties of the wall material. A further section presents a description of a computer program which uses the mathematical procedure described. The results of some test runs are followed by those of detailed parameter studies. In the course of these calculations the influences of a number of design and operational quantities of a fusion reactor were investigated. It turned out that the choice of wall thickness and wall loading are of predominant importance for the first wall fatigue life. (orig.) [de

Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity

International Nuclear Information System (INIS)

Fornos Herrando, J.

2013-01-01

The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

Falling liquid film flow along cascade-typed first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, T.; Nakai, T.; Kawara, Z.

2007-01-01

To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a 'cascade-typed' first wall with a falling liquid film flow is proposed as the 'liquid wall' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the STREAM code and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ρu 2 δ/σ: ρ is density, u is velocity, δ is film thickness, σ is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant water-head located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same structure and the same height as the reactor design

Flow Boiling on a Downward-Facing Inclined Plane Wall of Core Catcher

International Nuclear Information System (INIS)

Kim, Hyoung Tak; Bang, Kwang Hyun; Suh, Jung Soo

2013-01-01

In order to investigate boiling behavior on downward-facing inclined heated wall prior to the CHF condition, an experiment was carried out with 1.2 m long rectangular channel, inclined by 10 .deg. from the horizontal plane. High speed video images showed that the bubbles were sliding along the heated wall, continuing to grow and combining with the bubbles growing at their nucleation sites in the downstream. These large bubbles continued to slide along the heated wall and formed elongated slug bubbles. Under this slug bubble thin liquid film layer on the heated wall was observed and this liquid film prevents the wall from dryout. The length, velocity and frequency of slug bubbles sliding on the heated wall were measured as a function of wall heat flux and these parameters were used to develop wall boiling model for inclined, downward-facing heated wall. One approach to achieve coolable state of molten core in a PWR-like reactor cavity during a severe accident is to retain the core melt on a so-called core catcher residing on the reactor cavity floor after its relocation from the reactor pressure vessel. The core melt retained in the core catcher is cooled by water coolant flowing in an inclined cooling channel underneath as well as the water pool overlaid on the melt layer. Two-phase flow boiling with downward-facing heated wall such as this core catcher cooling channel has drawn a special attention because this orientation of heated wall may reach boiling crisis at lower heat flux than that of a vertical or upward-facing heated wall. Nishikawa and Fujita, Howard and Mudawar, Qiu and Dhir have conducted experiments to study the effect of heater orientation on boiling heat transfer and CHF. SULTAN experiment was conducted to study inclined large-scale structure coolability by water in boiling natural convection. In this paper, high-speed visualization of boiling behavior on downward-facing heated wall inclined by 10 .deg. is presented and wall boiling model for the

Numerical study of natural melt convection in cylindrical cavity with hot walls and cold bottom sink

Directory of Open Access Journals (Sweden)

Ahmanache Abdennacer

2013-01-01

Full Text Available Numerical study of natural convection heat transfer and fluid flow in cylindrical cavity with hot walls and cold sink is conducted. Calculations are performed in terms of the cavity aspect ratio, the heat exchanger length and the thermo physical properties expressed via the Prandtl number and the Rayleigh number. Results are presented in the form of isotherms, streamlines, average Nusselt number and average bulk temperature for a range of Rayleigh number up to 106. It is observed that Rayleigh number and heat exchanger length influences fluid flow and heat transfer, whereas the cavity aspect ratio has no significant effects.

Program system for calculating streaming neutron radiation field in reactor cavity

International Nuclear Information System (INIS)

He Zhongliang; Zhao Shu.

1986-01-01

The A23 neutron albedo data base based on Monte Carlo method well agrees with SAIL albedo data base. RSCAM program system, using Monte Carlo method with albedo approach, is used to calculate streaming neutron radiation field in reactor cavity and containment operating hall. The dose rate distributions calculated with RSCAM in square concrete duct well agree with experiments

The new electricity of France PWR: calculation scheme of neutron leakages from the reactor cavity

International Nuclear Information System (INIS)

Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.; Champion, G.

1987-04-01

A new calculation scheme is adapted to evaluate neutron fluxes in the reactor cavity and the containment of next french PWR. In this scheme a large part is given to Monte Carlo method, coupled with SN-method, in order to take into account multiple neutron diffusions and the complexity of the reactor geometry

Simulation of fusion first-wall environment in a fission reactor

International Nuclear Information System (INIS)

Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

1982-01-01

A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

Repetition rates in heavy ion beam driven fusion reactors

Science.gov (United States)

Peterson, Robert R.

1986-01-01

The limits on the cavity gas density required for beam propagation and condensation times for material vaporized by target explosions can determine the maximum repetition rate of Heavy Ion Beam (HIB) driven fusion reactors. If the ions are ballistically focused onto the target, the cavity gas must have a density below roughly 10-4 torr (3×1012 cm-3) at the time of propagation; other propagation schemes may allow densities as high as 1 torr or more. In some reactor designs, several kilograms of material may be vaporized off of the target chamber walls by the target generated x-rays, raising the average density in the cavity to 100 tor or more. A one-dimensional combined radiation hydrodynamics and vaporization and condensation computer code has been used to simulate the behavior of the vaporized material in the target chambers of HIB fusion reactors.

Repetition rates in heavy ion beam driven fusion reactors

International Nuclear Information System (INIS)

Peterson, R.R.

1986-01-01

The limits on the cavity gas density required for beam propagation and condensation times for material vaporized by target explosions can determine the maximum repetition rate of Heavy Ion Beam (HIB) driven fusion reactors. If the ions are ballistically focused onto the target, the cavity gas must have a density below roughly 10 -4 torr (3 x 10 12 cm -3 ) at the time of propagation; other propagation schemes may allow densities as high as 1 torr or more. In some reactor designs, several kilograms of material may be vaporized off of the target chamber walls by the target generated x-rays, raising the average density in the cavity to 100 tor or more. A one-dimensional combined radiation hydrodynamics and vaporization and condensation computer code has been used to simulate the behavior of the vaporized material in the target chambers of HIB fusion reactors

A SRF niobium cylindrical cavity with a large silicon nitride niobium-coated membrane as one end-wall

Science.gov (United States)

Martinez, Luis; Castelli, Alessandro; Pate, Jacob; Thompson, Johnathon; Delmas, William; Sharping, Jay; Chiao, Raymond; Chiao Team; Sharping Team

The development of large silicon nitride membranes and niobium film deposition techniques motivate new architectures in opto-mechanics and microwave devices that can exploit the extremely high Q's obtainable with superconducting radio frequency (SRF) niobium cavities. We present a X-band SRF cylindrical cavity-membrane system in which one end-wall of the cavity is replaced by a niobium coated centimeter-sized silicon nitride membrane. We report moderately high Q factors above 10 million. Experimental results characterizing the system and potential future applications for such schemes in microwave devices and optomechanics are discussed.

Engineering the fusion reactor first wall

International Nuclear Information System (INIS)

Wurden, Glen; Scott, Willms

2008-01-01

Recently the National Academy of Engineering published a set of Grand Challenges in Engineering in which the second item listed was entitled 'Provide energy from fusion'. Clearly a key component of this challenge is the science and technology associated with creating and maintaining burning plasmas. This is being vigorously addressed with both magnetic and inertial approaches with various experiments such as ITER and NIF. Considerably less attention is being given to another key component of this challenge, namely engineering the first wall that will contain the burning plasma. This is a daunting problem requiring technologies and materials that can not only survive, but also perform multiple essential functions in this extreme environment. These functions are (1) shield the remainder of the device from radiation. (2) convert of neutron energy to useful heat and (3) breed and extract tritium to maintain the reactor fuel supply. The first wall must not contaminate the plasma with impurities. It must be infused with cooling to maintain acceptable temperatures on plasma facing and structural components. It must not degrade. It must avoid excessive build-up of tritium on surfaces, and, if surface deposits do form, must be receptive to cleaning techniques. All these functions and constraints must be met while being subjected to nuclear and thermal radiation, particle bombardment, high magnetic fields, thermal cycling and occasional impingement of plasma on the surface. And, operating in a nuclear environment, the first wall must be fully maintainable by remotely-operated manipulators. Elements of the first wall challenge have been studied since the 1970' s both in the US and internationally. Considerable foundational work has been performed on plasma facing materials and breeding blanket/shield modules. Work has included neutronics, materials fabrication and joining, fluid flow, tritium breeding, tritium recovery and containment, energy conversion, materials damage and

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

International Nuclear Information System (INIS)

Conlin, J.A.

1985-08-01

A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

Study of liquid metal mixed convection in cavities

International Nuclear Information System (INIS)

Abadie, Philippe.

1979-10-01

This study has enabled some results to be obtained on the flow of liquid metals in cavities. The effects of different adimensional parameters characteristic of mixed convection flows were experimentally demonstrated. In the case of a roof heated cavity, three zones were distinguished: the mixing zone at the channel exit, a quasi constant temperature recirculation zone and a stratified zone at the top of the cavity. The thickness of this last region depends on natural convection effects: it disappears completely in a pure forced convection regime. A simple model using a critical Richardson number concept was developed in order to be able to predict the thickness of this region. Heat transfer correlation formulas were established both for the heated roof and forward direction heated wall cases. Some data was also obtained on temperature fluctuations for both cases. The different topics investigated are useful for defining heat transfers in certain regions of fast neutron sodium cooled reactors. A more extensive program is currently being developed in order to be able to investigate a wider range of variations in the above mentioned parameters and to more closely approximate reactor vessels [fr

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

First wall response to energy disposition in conceptual laser fusion reactors

International Nuclear Information System (INIS)

Hovingh, J.

1976-02-01

Discussed are energy depositions in the first wall of various proposed laser-fusion reactors and the effect of pulse time on the stress and temperature in the first wall. Simple models can be used to estimate the temperature and stress rise from x-rays and neutrons. More complex analysis is needed to estimate the response of the first wall to reflected laser light and the pellet debris

A model of gas cavity breakup behind a blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1980-05-01

A semi-empirical model has been developed to describe the transient behaviour of a gas cavity due to breakup behind a blockage in Liquid Metal Fast Breeder Reactor subassembly geometry. The main mechanisms assumed for gas cavity breakup in the present model are as follows: The gas cavity is broken up by the pressure fluctuation at the interface due to turbulence in the liquid. The centrifugal force on the liquid opposes breakup. The model is able to describe experimental results on the transient behaviour of a gas cavity due to breakup after the termination of gas injection. On the basis of the present model the residence time of a gas cavity behind a blockage in sodium is predicted and the dependence of the residence time on blockage size is discussed. (orig.) [de

Conception of thermonuclear reactor with a shielding layer of the first wall

International Nuclear Information System (INIS)

Marin, S.V.

1979-01-01

Considered is the way of the shielding of the first wall of a thermonuclear reactor by the layer of ISSEC (Internal spectral shifter and Energy Converter). It is a constructive non-power element placed between a plasma and the first wall, and intended for the softening of the spectrum and intensity reduction of particle fluxes falling on the first wall. Results of neutron-physical calculations of the UWMAK-type reactor blanket (in the S 4 -P 3 approximation) are presented. While comparing five materials (C, Mo, Nb, V,W) by the rate of radiation damage formation, gas production, radioactivity level and energy output in the blanket with the 316 stainless steel first wall, it is obvious that the conception of ISSEC permits to prolong the service period of the first wall. Construction elements should be then in the same irradiation conditions as those in fast reactors. Molybdenum has been taken as the best ISSEC material. It reduces the number of displaced atoms of the first wall by 20% and decreases helium production by about 100%, increases energy output in the blanket by 15-18%. However, graphite is advantageous, while comparing it to molybdenum in values of residual energy output, radioactivity level, costs and manufacture simplicity. One problem stays unsolved, which is connected with chemical sputtering of graphite at the formation of C 2 H 2 in the high temperature range. So it is hard to prefer any material now

SOLASE conceptual laser fusion reactor study

International Nuclear Information System (INIS)

Moses, G.A.; Conn, R.W.; Abdel-Khalik, S.I.; Cooper, G.W.; Howard, J.; Magelssen, G.R.

1978-01-01

A conceptual laser fusion reactor for electric power, SOLASE, has been designed. The SOLASE design utilizes a 1 MJ, 6.7% efficient laser to implode 20 fusion targets per second. The target gain is 150 and produces a net electrical power of 1000 MW. The reactor cavity is spherical with a 6 m radius. The first wall is graphite and has a neutron wall loading of 5 MW/m 2 . It is protected from the target debris by low pressure xenon gas that is introduced into the cavity. The blanket structure is a honeycombed graphite composite. The tritium breeding and heat transport medium is Li 2 O in the form of pellets that flow through the blanket. The tritium breeding ration is 1.34. Temperature decoupling of the graphite structure and the Li 2 O coolant enables the structure to operate at temperatures that minimize radiation damage effects. The graphite blanket is replaced every year but exhibits low levels of radioactivity so that limited hands on maintenance is possible two weeks after shutdown, thus facilitating rapid replacement

Production cavity and central optics for a light shining through a wall experiment

International Nuclear Information System (INIS)

Hodajerdi, Reza

2015-02-01

The unexplained nature of dark matter and dark energy is a prominent reason for investigating physics beyond the standard model of particle physics (SM). Some extensions of the SM propose weakly interacting slim particles (WISPs). In an attempt to prove the existence of these particles, Light shining through the wall (LSW) experiments explore a very weak coupling between WISPs and photons (and viceversa). LSW experiments employ high-power lasers that provide a well defined flux of photons for the WISP-Photon conversion. The ALPS-I experiment at DESY in Hamburg was the first successful experiment with a high finesse optical resonator to enhance the laser power in a strong magnetic field in order to increase the photon to WISP conversion probability. The ALPS-II experimental concept adds a second optical cavity to also increase the reconversion probability. Both cavities are separated by a wall, amplify light at 1064 nm and share a common optical axis. Operating these two cavities inside 20 straightened HERA superconducting dipole magnets and using a transition edge sensor (TES) as a single photon detector will make the ALPS-II experiment almost three orders of magnitude more sensitive than its predecessor. Since photons, originating from reconverted WISPs in the regeneration cavity (RC) have 1064 nm wavelengths, the RC has to be locked to the production cavity (PC) with light of a different wavelength. Therefore frequency doubled PCs light will be used to lock the RC. This 532 nm light shall not arrive at the TES to prevent background noise. To achieve this, an optical attenuation system for wavelengths different from 1064 nm is required. In my thesis, the required attenuation was estimated and an optical setup was proposed and constructed and tested. It attenuates green photons by a factor of of 10 -18 and transmits 85% of the infrared photons. Furthermore the high finesse production cavity of ALPS-IIa was set up and characterized during this thesis. The PC reached

Optical cavity furnace for semiconductor wafer processing

Science.gov (United States)

Sopori, Bhushan L.

2014-08-05

An optical cavity furnace 10 having multiple optical energy sources 12 associated with an optical cavity 18 of the furnace. The multiple optical energy sources 12 may be lamps or other devices suitable for producing an appropriate level of optical energy. The optical cavity furnace 10 may also include one or more reflectors 14 and one or more walls 16 associated with the optical energy sources 12 such that the reflectors 14 and walls 16 define the optical cavity 18. The walls 16 may have any desired configuration or shape to enhance operation of the furnace as an optical cavity 18. The optical energy sources 12 may be positioned at any location with respect to the reflectors 14 and walls defining the optical cavity. The optical cavity furnace 10 may further include a semiconductor wafer transport system 22 for transporting one or more semiconductor wafers 20 through the optical cavity.

Resistive requirements for the vacuum wall of a tokamak fusion reactor

International Nuclear Information System (INIS)

Brooks, J.N.; Ehat, D.; Harkness, S.D.; Norem, J.; Stevens, H.; Turner, L.

1978-01-01

Most conceptual designs of tokamak power reactors have incorporated a ceramic insulator in the vacuum wall to make the wall electrically non-conducting. Such a material will have to be highly resistant to radiation damage at doses up to at least 10 MW-yr/m 2 while being compatible with a coolant and a first wall whose dimensions change due to thermal cycling and radiation damage. Thus there is considerable incentive to assess the consequences of eliminating the flux breaker from the design and having a conducting boundary instead. In this initial study the question of having a finite wall resistance has been examined in terms of its major implications on both the normal and abnormal operation of a tokamak reactor. This study has been conducted within the framework of the ANL-EPR-77 design although the results should provide some guidance for future reactors as well. The EPR design referred to is a 5 m major radius tokamak with an aspect ratio of 3.5, and with an equilibrium plasma current of 7.3 MA. The vacuum chamber is designed to accommodate a non-circular plasma with a height to width ratio of up to 1.65. The basic vacuum wall design is shown in Fig. 1. It is located about 0.4 M from the plasma boundary and has an irregular polygon shape made of sixteen sections, one per TF coil interval. Variations of this design having a range of resistance values have been used in the analysis

Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

Energy Technology Data Exchange (ETDEWEB)

Lafitte, R.; Marchand, J. D. [Bonnard et Gardel, Ingenieurs-Conseil, Lausanne (Switzerland)

1981-01-15

The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed.

Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

International Nuclear Information System (INIS)

Lafitte, R.; Marchand, J.D.

1981-01-01

The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed

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

Gas core reactor power plants designed for low proliferation potential

International Nuclear Information System (INIS)

Lowry, L.L.

1977-09-01

The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

In-Vessel Retention via External Reactor Cooling

Energy Technology Data Exchange (ETDEWEB)

Bachrata, Andrea [CTU in Prague, Faculty of nuclear sciences and physical engineering, V Holesovickach 2 180 00, Prague 8 (Czech republic)

2008-07-01

In-vessel (corium) retention (IVR) via external reactor pressure vessel (RPV) cooling is considered to be an effective severe accident management strategy for corium localisation and stabilisation. The main idea of IVR strategy consists in flooding the reactor cavity and transferring the decay heat through the wall of RPV to the recirculating water and than to the atmosphere of the containment of nuclear power plant. The aim of this strategy is to localise and to stabilise the corium inside the RPV. Not using this procedure could destroy the integrity of RPV and might cause the interaction of the corium with the concrete at the bed of the reactor cavity. Several experimental facilities and computer codes (MVITA, ASTEC module DIVA and CFD codes) were applied to simulate the IVR strategy for concrete reactor designs. The necessary technical modifications concerning the implementation of IVR concept were applied at the Loviisa NPP (VVER-440/V213). This strategy is also an important part of the advanced reactor designs AP600 and AP1000. (authors)

Cavity structural integrity evaluation of steam explosion using LS-DYNA

Energy Technology Data Exchange (ETDEWEB)

Lee, Dae-Young; Park, Chang-Hwan [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Kim, Kap-sun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

For investigating the mechanical response of the newly-designed NPP against an steam explosion, the cavity structural integrity evaluation was performed, in which the mechanical load resulted from a steam explosion in the reactor cavity was calculated. In the evaluation, two kinds of approach were considered, one of which is a deterministic manner and the other is a probabilistic one. In this report, the procedure and the results of the deterministic analysis are presented When entering the severe accident, the core is relocated to the lower head. In this case, an Ex-Vessel Steam Explosion(EVSE) can occur. It can threaten the structural integrity of the cavity due to the load applied to the walls or slabs of the cavity. The large amount of the energy transmitted from interaction between the molten corium and the water causes a dynamic loading onto the concrete walls resulting not only to affect the survivability of the various equipment but also to threaten the integrity of the containment. In this report, the response of the cavity wall structure is analyzed using the nonlinear finite element analysis (FEA) code. The resulting stress and strain of the structure were evaluated by the criteria in NEI07-13. Until now, deterministic analysis was performed via finite element analysis for the dynamic load generated by the steam explosion to investigate the effect on the cavity structure. A deterministic method was used in this study using the specific values of material properties and clearly defined steam explosion pressure curve. The results showed that the rebar and the liner are kept intact even at the high pressure pulse given by the steam explosion. The liner integrity is more critical to judge the preservation of the lean-tightness. In the meantime, there were found cracks in concrete media.

A design of a first wall for a demo reactor

International Nuclear Information System (INIS)

Bond, A.; Bond, R.A.; Cooke, P.I.H.

1985-01-01

A design of a first wall for a Demonstration reactor is reported based on an analysis of heat trasnport, sputtering damage, blanket neutronics and vacuum characteristics. The design comprises replaceable tungsten tiles radiatively cooled to a copper substrate, which in turn is cooled by high pressure helium. The overall engineering design of the first wall is described together with a discussion of the factors influencing the choice of design and materials

Circulation system for flowing uranium hexafluoride cavity reactor experiments

International Nuclear Information System (INIS)

Jaminet, J.F.; Kendall, J.S.

1976-01-01

Accomplishment of the UF 6 critical cavity experiments, currently in progress, and planned confined flowing UF 6 initial experiments requires development of reliable techniques for handling heated UF 6 throughout extended ranges of temperature, pressure, and flow rate. The development of three laboratory-scale flow systems for handling gaseous UF 6 at temperatures up to 500 K, pressures up to approximately 40 atm, and continuous flow rates up to approximately 50 g/s is presented. A UF 6 handling system fabricated for static critical tests currently being conducted at Los Alamos Scientific Laboratory (LASL) is described. The system was designed to supply UF 6 to a double-walled aluminum core canister assembly at temperatures between 300 K and 400 K and pressures up to 4 atm. A second UF 6 handling system designed to provide a circulating flow of up to 50 g/s of gaseous UF 6 in a closed-loop through a double-walled aluminum core canister with controlled temperature and pressure is described

High-flux first-wall design for a small reversed-field pinch reactor

International Nuclear Information System (INIS)

Cort, G.E.; Graham, A.L.; Christensen, K.E.

1982-01-01

To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high-technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time-averaged heat flux of 4.5 MW/m 2 with a conservatively estimated transient peak approximately twice the average value. We present the design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities. We combined a thermal analysis with a structural fatigue analysis to design the heat transfer module to last 10 6 cycles or one year at 80% duty for a 26-s power cycle. This fatigue life is compatible with a radiation damage life of 14 MW/yr/m 2

Ultimate shearing strength of aseismatic walls with many small holes for reactor buildings

International Nuclear Information System (INIS)

Yoshizaki, Seiji; Ezaki, Tetsuro; Korenaga, Takeyoshi; Sotomura, Kentaro.

1984-01-01

The aseismatic walls for reactor buildings have complicated forms, and are characterized by large wall thickness and high reinforcement ratio as compared with ordinary aseismatic walls. The forms are mainly box, cylinder or irregular polygonal prism and their combination. The design of the walls with many small holes has been performed on the basis of the reinforced concrete structure calculation standard of the Architectural Institute of Japan, following the case with large opening. When there are many small holes, the arrangement of reinforcement for the openings becomes complex, and the construction is difficult. It is necessary to rationalize the design and to simplify the reinforcement work. Under the background like this, the experiment to examine the shearing property in bending of the aseismatic walls with many small holes for reactor buildings was carried out, and horizontal loading test was performed on 43 specimens. The method of calculating the ultimate shearing strength of a wall without opening was proposed, and the method of applying it to a wall with many small holes is shown. The experimental method and the results, the examination of the experimental results, and the ultimate shearing strength of the aseismatic walls are reported. (Kako, I.)

Shear flow over a plane wall with an axisymmetric cavity or a circular orifice of finite thickness

International Nuclear Information System (INIS)

Pozrikidis, C.

1994-01-01

Shear flow over a plane wall that contains an axisymmetric depression or pore is studied using a new boundary integral method which is suitable for computing three-dimensional Stokes flow within axisymmetric domains. Numerical results are presented for cavities in the shape of a section of a sphere or a circular cylinder of finite length, and for a family of pores or orifices with finite thickness. The results illustrate the distribution of shear stresses over the plane wall and inside the cavities or pores. It is found that in most cases, the distribution of shear stresses over the plane wall, around the depressions, is well approximated with that for flow over an orifice of infinitesimal thickness for which an exact solution is available. The kinematic structure of the flow is discussed with reference to eddy formation and three-dimensional flow reversal. It is shown that the thickness of a circular orifice or depth of a pore play an important role in determining the kinematical structure of the flow underneath the orifice in the lower half-space

Dynamic loading of the structural wall in a lithium fall fusion reactor

International Nuclear Information System (INIS)

Glenn, L.A.

1979-01-01

In one version of an inertial confinement fusion (ICF) power reactor, the laser-imploded pellet is surrounded by a thick, annular 'waterfall' of liquid lithium. The fall has three functions: to breed tritium for pellet resupply, to act as an energy sink and heat exchange mdeium with an external power loop, and to protect the first wall of the reactor from excessive neutronic and hydrodynamic loading. Our primary concern here is with this last function. We formulated a simple model of a lithium-fall ICF reactor and calculated the fall disassembly and the subsequent fluid-wall interaction resulting from the energy deposition by the imploded pellet. Two potential mechanisms for wall damage were identified: surface erosion and hoop failure. For single fall designs, the erosion problem appears to be serious. Concentric annuli (multiple fall) or packed jet configurations may be feasible but experiments are needed to clarify the physical model, especially with reg (orig.)ard to /orig.the characteristics of the cavitated liquid lithium and of the two-phase liquid-vapor region.

Review of melting and evaporation of fusion-reactor first walls

International Nuclear Information System (INIS)

Fillo, J.A.; Makowitz, H.

1981-01-01

The most severe thermal loading on the first wall will occur when the plasma becomes unstable resulting in a hard plasma disruption or at the end of a discharge when the plasma is dumped on the wall in a very short period of time. Hard plasma disruptions are of particular concern in future fusion reactors where the thermal energy of the plasma may reach values on the order of 300 MJ. Sufficiently high heating rates can occur to melt the first wall surface, and the temperature can increase resulting in vaporization. Thermal models are reviewed which treat these problems

Modeling of heat transfer in wall-cooled tubular reactors

NARCIS (Netherlands)

Koning, G.W.; Westerterp, K.R.

1999-01-01

In a pilot scale wall-cooled tubular reactor, temperature profiles have been measured with and without reaction. As a model reaction oxidation of carbon monoxide in air over a copper chromite catalyst has been used. The kinetics of this reaction have been determined separately in two kinetic

CT findings of solitary tuberculoma with a cavity

Energy Technology Data Exchange (ETDEWEB)

Goo, Dong Erk; Goo, Hyun Woo; Song, Koun Sik; Lim, Tae Hwan; Kim, Won Dong [Asan Medical Center, University of Ulsan College of Medicine, Seoul (Korea, Republic of)

1994-09-15

Differential diagnosis of solitary pulmonary nodule with cavity includes lung abscess, tuberculoma, bronchogenic carcinoma, metastasis and trauma, etc. We analyzed the CT appearance of tuberculoma presenting as a solitary pulmonary nodule with cavity and describe the findings which suggest tuberculoma in the differential diagnosis of solitary pulmonary nodule with cavity. 25 patients with solitary pulmonary nodule(diameter less than 4 cm) without surrounding parenchymal consolidation on chest radiograph, who had a cavity within the nodule on CT, were included in our study. Density of the nodule, maximal wall thickness, the character of inner and outer wall margin, location of cavity within nodule, location of the nodule, presence or absence of satellite lesions and calcification were analyzed. Solitary tuberculoma with cavity showed maximal wall thickness more than 15 m in 40%(10/25) and 5-14 mm in 56%(14/25), eccentric cavitation in 84%(21/25) and concentric cavitation in 16%(4/25), spiculated outer wall margin in 56%(14/15) and lobulated margin in 32%(8/25), smooth inner wall margin in 60%(15/25) and nodular margin in 40%(10/25). CT density of the cavity wall compared wth the chest wall muscle was low in 84%(21/25) and isodense in 16%(4/25). Accompanying satellite lesions were seen in 84%(21/25) and calcification was visible in 28%(7/25). The CT findings of solitary tuberculoma with cavity are relative peripheral location, eccentric cavitation, finely spiculated outer wall margin, and mean maximal wall thickness of 13.2 mm, which are also the common features of malignant nodule. However, relative low density of the nodule compared to the chest wall muscle and surrounding satellite lesions can be additional clues favouring solitary tuberculoma with cavity on CT.

First wall lifetime of the near term fusion reactors

International Nuclear Information System (INIS)

Matera, R.; Botti, S.; Cerrai, G.

1985-01-01

A sensitivity analysis of the influence of the operating conditions and of the design parameters over the first wall lifetime was performed by means of the computer program smile. In the range of operating conditions typical of an experimental fusion reactor like NET/INTOR and for a type AISI 316 stainless steel structural material, fatigue damage and fatigue crack growth are the limiting failure mechanisms of the first wall. The analysis shows in graphical form the limits of the allowable range of operating conditions or of design parameters

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

Science.gov (United States)

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

An experimental study of flame stability in a directly-fueled wall cavity with a supersonic free stream

Science.gov (United States)

Rasmussen, Chadwick Clifford

An extensive study of flame stability in a cavity-based fuel injector/flameholder has been performed. Flames were stabilized in cavities with two different aft wall configurations and length to depth ratios of 3 and 4. Fuel was injected directly into the cavity using two injector configurations. Fuel injected from the aft wall of the cavity entered directly into the recirculation zone and provided desirable performance near the lean blowout limit. At high fuel flowrates, the cavity became flooded with fuel and rich blowout occurred. When fuel was injected from the floor of the cavity, excess fuel was directed out of the cavity which allowed for flame stabilization at extremely high fuel flowrates; however, this phenomenon also resulted in suboptimal performance near the lean limit where the blowout point was less predictable. Images of planar laser-induced fluorescence (PLIF) of CH, OH, and formaldehyde give insight into the flameholding mechanisms. CH layers in the cavity are thin and continuous and show structure that is comparable to lifted jet flames, while broad CH zones are sometimes observed in the shear layer. OH PLIF images show that hot recirculated products are always present at the location of flame stabilization, whereas images of formaldehyde indicate that partial premixing takes place in the shear layer portion of the flame. Nonreacting measurements of the boundary layer and the free stream velocity profiles were obtained to provide necessary boundary conditions for computational modeling. Mean and instantaneous velocity profiles were determined for the nonreacting flow using particle image velocimetry (PIV). A correlation of the blowout points for a directly-fueled cavity in a supersonic flow was accomplished using a Damkohler number and an equivalence ratio based upon an effective air mass flowrate. The chemical time was formulated using a generic measure of the reaction rate, tauc ˜ alpha/ S2L , which was found to be adequate for correlating lean

Gaseous-fuel nuclear reactor research for multimegawatt power in space

Science.gov (United States)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Materials for heat flux components of the first wall in fusion reactors

International Nuclear Information System (INIS)

Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1985-08-01

Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

Dependence of the Casimir-Polder interaction between an atom and a cavity wall on atomic and material properties

International Nuclear Information System (INIS)

Mostepanenko, V M; Babb, J F; Caride, A O; Klimchitskaya, G L; Zanette, S I

2006-01-01

The Casimir-Polder and van der Waals interactions between an atom and a flat cavity wall are investigated under the influence of real conditions including the dynamic polarizability of the atom, actual conductivity of the wall material and nonzero temperature of the wall. The cases of different atoms near metal and dielectric walls are considered. It is shown that to obtain accurate results for the atom-wall interaction at short separations, one should use the complete tabulated optical data for the complex refractive index of the wall material and the accurate dynamic polarizability of an atom. At relatively large separations in the case of a metal wall, one may use the plasma model dielectric function to describe the dielectric properties of the wall material. The obtained results are important for the theoretical interpretation of experiments on quantum reflection and Bose-Einstein condensation

Cavity pressure history of contained nuclear explosions

Energy Technology Data Exchange (ETDEWEB)

Chapin, C E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

1970-05-01

Knowledge of pressure in cavities created by contained nuclear explosions is useful for estimating the possibility of venting radioactive debris to the atmosphere. Measurements of cavity pressure, or temperature, would be helpful in evaluating the correctness of present code predictions of underground explosions. In instrumenting and interpreting such measurements it is necessary to have good theoretical estimates of cavity pressures. In this paper cavity pressure is estimated at the time when cavity growth is complete. Its subsequent decrease due to heat loss from the cavity to the surrounding media is also predicted. The starting pressure (the pressure at the end of cavity growth) is obtained by adiabatic expansion to the final cavity size of the vaporized rock gas sphere created by the explosion. Estimates of cavity size can be obtained by stress propagation computer codes, such as SOC and TENSOR. However, such estimates require considerable time and effort. In this paper, cavity size is estimated using a scheme involving simple hand calculations. The prediction is complicated by uncertainties in the knowledge of silica water system chemistry and a lack of information concerning possible blowoff of wall material during cavity growth. If wall material blows off, it can significantly change the water content in the cavity, compared to the water content in the ambient media. After cavity growth is complete, the pressure will change because of heat loss to the surrounding media. Heat transfer by convection, radiation and conduction is considered, and its effect on the pressure is calculated. Analysis of cavity heat transfer is made difficult by the complex nature of processes which occur at the wall where melting, vaporization and condensation of the gaseous rock can all occur. Furthermore, the melted wall material could be removed by flowing or dripping to the cavity floor. It could also be removed by expansion of the steam contained in the melt (blowoff) and by

A three-bar model for ratcheting of fusion reactor first wall

International Nuclear Information System (INIS)

Wolters, J.; Majumdar, S.

1994-12-01

First wall structures of fusion reactors are subjected to cyclic bending stresses caused by inhomogeneous temperature distribution during plasma burn cycles and by electromagnetically induced impact loads during plasma disruptions. Such a combination of loading can potentially lead to ratcheting or incremental accumulation of plastic strain with cycles. An elastic-plastic three-bar model is developed to investigate the ratcheting behavior of the first wall

Transient temperature and stress distributions in the pressure vessel's wall of a nuclear reactor

International Nuclear Information System (INIS)

Silva, G.A. da

1979-01-01

In order to calculate the temperature distribution in a reactor vessel wall which is under the effect of gamma radiation originated in the reactor core, a numerical solution is proposed. This problem may arise from a reactor cooling pump failure .The thermal stresses are also calculated. (Author) [pt

Investigation of cascade-typed falling liquid film flow along first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, Tomoaki; Nakai, Tadakatsu; Kawara, Zensaku

2007-01-01

To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a ''cascade-typed'' falling liquid film flow is proposed as the ''liquid wall'' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the commercial code (STREAM: unsteady three-dimensional general purpose thermofluid code) and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ru 2 d/s: r is density, u is velocity, d is film thickness, s is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant waterhead located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same

RF Behavior of Cylindrical Cavity Based 240 GHz, 1 MW Gyrotron for Future Tokamak System

Science.gov (United States)

Kumar, Nitin; Singh, Udaybir; Bera, Anirban; Sinha, A. K.

2017-11-01

In this paper, we present the RF behavior of conventional cylindrical interaction cavity for 240 GHz, 1 MW gyrotron for futuristic plasma fusion reactors. Very high-order TE mode is searched for this gyrotron to minimize the Ohmic wall loading at the interaction cavity. The mode selection process is carried out rigorously to analyze the mode competition and design feasibility. The cold cavity analysis and beam-wave interaction computation are carried out to finalize the cavity design. The detail parametric analyses for interaction cavity are performed in terms of mode stability, interaction efficiency and frequency. In addition, the design of triode type magnetron injection gun is also discussed. The electron beam parameters such as velocity ratio and velocity spread are optimized as per the requirement at interaction cavity. The design studies presented here confirm the realization of CW, 1 MW power at 240 GHz frequency at TE46,17 mode.

Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

International Nuclear Information System (INIS)

Toshimas, M.; Nobuo, S.

1983-01-01

For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

Liquid wall boiler and moderator (BAM) for heavy ion-pellet fusion reactors

International Nuclear Information System (INIS)

Powell, J.R.; Lazareth, O.; Fillo, J.

1977-11-01

Thick liquid wall blankets appear to be of great promise for heavy ion pellet fusion reactors. They avoid the severe problems of intense radiation and blast damage that would be encountered with solid blanket structures. The liquid wall material can be chosen so that its vapor pressure at the working temperature of the power cycle is well below the value at which it might interfere with the propagation of the heavy ion beam. The liquid wall can be arranged so that it does not contact any surrounding solid structure when the pellet explosion occurs, including the ends. The ends can be magnetically closed just before the pellet explosion, or a time phased flow can be used, which will leave a clear central zone into which the pellet is injected. Parametric analysis comparing three candidate liquid wall materials were carried out. The three materials were lithium, flibe, and lead (with a low concentration of disolved lithium). Lead appeared to be the best choice for the liquid wall, although any of the three should allow a practical reactor system. The parametric analyses examined the effects of pellet yield (0 to 10 GJ), pellet mass (3 g to 3 kg), liquid wall thickness (10 cm to 80 cm), vapor condensation time (0 to 10 milliseconds), degree of neutron moderation in the pellet (none to 100%), liquid wall chamber size (radius of 1.5 meters to 4 meters), Pb/Li 6 ratio (100 to 5,000), and thickness of graphite moderating zone behind the liquid wall

Thermal responses of tokamak reactor first walls during cyclic plasma burns

International Nuclear Information System (INIS)

Smith, D.L.; Charak, I.

1978-01-01

The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer code has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wall and liner have been determined for several burn-cycle scenarios and the extent of surface ablation that results from a plasma disruption has been determined for stainless steel and graphite first surfaces

Thermal responses of tokamak reactor first walls during cyclic plasma burns

International Nuclear Information System (INIS)

Smith, D.L.; Charak, I.

1977-01-01

The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wall and liner have been determined for several burn-cycle scenarios and the extent of surface ablation that results from a plasma disruption has been determined for stainless steel and graphite first surfaces

Laser fusion reactor design in a fast ignition with a dry wall chamber

International Nuclear Information System (INIS)

Ogawa, Yichi; Goto, Takuya; Ninomiya, Daisuke; Hiwatari, Ryoji; Asaoka, Yoshiyuki; Okano, Kunihiko

2007-01-01

One of the critical issues in laser fusion reactor design is high pulse heat load on the first wall by the X-rays and the fast/debris ions from fusion burn. There are mainly two concepts for the first wall of laser fusion reactor, a dry wall and a liquid metal wall. We should notice that the fast ignition method can achieve sufficiently high pellet gain with smaller (about 1/10 of the conventional central ignition method) input energy. To take advantage of this property, the design of a laser fusion reactor with a small size dry wall chamber may become possible. Since a small fusion pulse leads to a small electric power, high repetition of laser irradiation is required to keep sufficient electric power. Then we tried to design a laser fusion reactor with a dry wall chamber and a high repetition laser. This is a new challenging path to realize a laser fusion plant. Based on the point model of the core plasma, we have estimated that fusion energy in one pulse can be reduced to be 40 MJ with a pellet gain around G>100. To evaluate the validity of this simple estimation and to optimize the pellet design and the pulse shaping for the fast ignition scenario, we have introduced 1-D hydrodynamic simulation code ILESTA-1D and carried out implosion simulations. Since the code is one-dimensional, the detailed physics process of fast heating cannot be reproduced. Thus the fast heating is reflected in the code as the additional artificial heating source in the energy equation. It is modeled as a homogeneous heating of electrons in core region at the time just before when the maximum compression is achieved. At present we obtained the pellet gain G∝100 with the same input energy as the above estimation by a simple point model (350kJ for implosion, 50kJ for heating and assuming 20% coupling of heating laser). A dry wall is exposed to several threats due to the cyclic load by the high energy X-ray and charged particles: surface melting, physical and chemical sputtering

A conceptual design strategy for liquid-metal-wall inertial fusion reactors

International Nuclear Information System (INIS)

Monsler, M.J.; Meier, W.R.

1981-01-01

The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade are reviewed from the perspective of formulating a conceptual design strategy for such chambers. The basis for the design strategy is set by enumerating both the attractive and unattractive features of a LMW chamber. Past concepts are then reviewed to identify conceptual design approaches and physical configurations that enhance the positive aspects and minimize the negative aspects. A detailed description of the engineering considerations is given, including such topics as the selection of a liquid metal, control of radiation damage, selection of structural material, control of tritium breeding and extraction, control of wall stress, and designing for a given rep-rate. Finally, a design strategy is formulated which accomodates the engineering constraints while minimizing the liquid-metal flow rate. (orig.)

Nuclear reactor buildings

International Nuclear Information System (INIS)

Nagashima, Shoji; Kato, Ryoichi.

1985-01-01

Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

International Nuclear Information System (INIS)

Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

1988-09-01

High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Richinson, P.J.

1978-08-01

The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

Assessment of models for steam release from concrete and implications for modeling corium behavior in reactor cavities

International Nuclear Information System (INIS)

Washington, K.E.; Carroll, D.E.

1988-01-01

Models for concrete outgassing have been developed and incorporated into a developmental version of the CONTAIN code for the assessment of corium behavior in reactor cavities. The resultant code, referred to as CONTAIN/OR in order to distinguish it from the released version of CONTAIN, has the capability to model transient heat conduction and concrete outgassing in core-concrete interaction problems. This study focused on validation and assessment of the outgassing model through comparisons with other concrete response codes. In general, the model is not mechanistic; however, there are certain important processes and feedback effects that are treated rigorously. The CONTAIN outgassing model was compared against two mechanistic concrete response codes (USINT and SLAM). Gas release and temperature profile predictions for several concrete thicknesses and heating rates were performed with acceptable agreement seen in each case. The model was also applied to predict corium behavior in a reactor cavity for a hypothetical severe accident scenario. In this calculation, gases evolving from the concrete during nonablating periods fueled exothermic Zr chemical reactions in the corium. Higher corium temperatures and more concrete ablation were observed when compared with that seen when concrete outgassing was neglected. Even though this result depends somewhat upon the makeup of the corium sources and the concrete type in the cavity, it does show that concrete outgassing can be important in the modeling of corium behavior in reactor cavities. In particular, the need to expand the traditional role of CORCON from steady-state ablation to the consideration of more transient events is clearly evident as a result of this work. 5 refs., 11 figs., 1 tab

Experimental study on joint construction method for aseismatic walls of reactor buildings, (1)

International Nuclear Information System (INIS)

Sugita, Kazunao; Mogami, Tatsuo; Ezaki, Tetsuro

1987-01-01

On the aseismatic walls of a reactor auxiliary building, many temporary openings are provided at the time of the construction for carrying equipment in later, due to the demand of shortening the construction period. Thus on the aseismatic walls, in most cases there are the joints due to the concrete placed later. As equipment tends to be unitized and become large, the quipment is placed close to the wall having an opening, consequently, the workability is poor, and the standardization of construction method is urgently demanded. The conventional method of closing temporary openings has the problems of safety and connecting reinforcing bars, therefore, the new construction method was proposed. In reactor buildings, the joints of walls are unavoidable, and since those are large scale structures, the joints are numerous. Therefore, at the joint parts, it abandoned and buried frames are used, it is advantageous in the time and cost of joint construction. In both cases, the mechanical properties were confirmed by the fundamental performance test partially modeling the joints and the verifying test modeling the whole walls. In this paper, the test of applying only shearing force to joint models is reported. (Kako, I.)

Plasma processing of superconducting radio frequency cavities

Science.gov (United States)

Upadhyay, Janardan

The development of plasma processing technology of superconducting radio frequency (SRF) cavities not only provides a chemical free and less expensive processing method, but also opens up the possibility for controlled modification of the inner surfaces of the cavity for better superconducting properties. The research was focused on the transition of plasma etching from two dimensional flat surfaces to inner surfaces of three dimensional (3D) structures. The results could be applicable to a variety of inner surfaces of 3D structures other than SRF cavities. Understanding the Ar/Cl2 plasma etching mechanism is crucial for achieving the desired modification of Nb SRF cavities. In the process of developing plasma etching technology, an apparatus was built and a method was developed to plasma etch a single cell Pill Box cavity. The plasma characterization was done with the help of optical emission spectroscopy. The Nb etch rate at various points of this cavity was measured before processing the SRF cavity. Cylindrical ring-type samples of Nb placed on the inner surface of the outer wall were used to measure the dependence of the process parameters on plasma etching. The measured etch rate dependence on the pressure, rf power, dc bias, temperature, Cl2 concentration and diameter of the inner electrode was determined. The etch rate mechanism was studied by varying the temperature of the outer wall, the dc bias on the inner electrode and gas conditions. In a coaxial plasma reactor, uniform plasma etching along the cylindrical structure is a challenging task due to depletion of the active radicals along the gas flow direction. The dependence of etch rate uniformity along the cylindrical axis was determined as a function of process parameters. The formation of dc self-biases due to surface area asymmetry in this type of plasma and its variation on the pressure, rf power and gas composition was measured. Enhancing the surface area of the inner electrode to reduce the

Comparative thermal performance of static sunshade and brick cavity wall for energy efficient building envelope in composite climate

Directory of Open Access Journals (Sweden)

Charde Meghana

2014-01-01

Full Text Available Energy efficient building technologies can reduce energy consumption in buildings. In present paper effect of designed static sunshade, brick cavity wall with brick projections and their combined effect on indoor air temperature has been analyzed by constructing three test rooms each of habitable dimensions (3.0 m × 4.0 m × 3.0 m and studying hourly temperatures on typical days for one month in summer and winter each. The three rooms have also been simulated using a software and the results have been compared with the experimental results. Designed static sunshade increased indoor air temperature in winter while proposed brick cavity wall with brick projections lowered it in summer. Combined effect of building elements lowered indoor air temperature in summer and increased it in winter as compared to outdoor air temperature. It is thus useful for energy conservation in buildings in composite climate.

Oxidation of ethene in a wall-cooled packed-bed reactor

NARCIS (Netherlands)

Schouten, E.P.S.; Borman, P.C.; Westerterp, K.R.

1994-01-01

The selective oxidation of ethene over a silver on α-alumina catalyst was studied in a pilot plant with a wall-cooled tubular packed bed reactor. Gas and solid temperatures in the catalyst bed were measured at different axial and radial positions as well as concentrations at different axial

Refractory oxides for fusion reactor first walls, the effects of the reducing environment

International Nuclear Information System (INIS)

Hoffman, J.G.

1979-01-01

Of the several applications for refractory oxides in fusion reactor systems, the most demanding is that for the first wall. Some components in proximity of the first wall (possibly waveguides or flux breakers) will also be subjected to similar environments. Many parameters affect the ultimate usability of a particular material for reactor applications: electrical resistivity and dielectric breakdown if applicable, thermal conductivity, mechanical properties, and stability with respect to neutral molecular or atomic, or ionized fuel gases. All these properties can be affected by the radiation environment present in an operating power reactor. Temperatures up to 2000K may be expected for radiatively cooled first wall liners in some proposed designs although surface temperatures are appreciably lower (approximately 1000K) in other applications. The exact nature of the chemical environment is not defined even for the most well developed design concepts, but possible environments may be hypothesized; ambient neutral molecular and atomic species, bombardment by high energy charge exchange neutral atoms, direct ionic bombardment from stray ions, and plasma dumps from failure of the confinement system. Preliminary work has begun to more adequately define the extent of the problem and suggest approaches to engineering solutions

New Measurements and Calculations to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

Energy Technology Data Exchange (ETDEWEB)

Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)

2011-07-01

Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)

Evolution of titanium residue on the walls of a plasma-etching reactor and its effect on the polysilicon etching rate

Energy Technology Data Exchange (ETDEWEB)

Hirota, Kosa, E-mail: hirota-kousa@sme.hitachi-hitec.com; Itabashi, Naoshi; Tanaka, Junichi [Hitachi, Ltd., Central Research Laboratory, 1-280, Higashi-Koigakubo, Kokubunji, Tokyo 185-8601 (Japan)

2014-11-01

The variation in polysilicon plasma etching rates caused by Ti residue on the reactor walls was investigated. The amount of Ti residue was measured using attenuated total reflection Fourier transform infrared spectroscopy with the HgCdTe (MCT) detector installed on the side of the reactor. As the amount of Ti residue increased, the number of fluorine radicals and the polysilicon etching rate increased. However, a maximum limit in the etching rate was observed. A mechanism of rate variation was proposed, whereby F radical consumption on the quartz reactor wall is suppressed by the Ti residue. The authors also investigated a plasma-cleaning method for the removal of Ti residue without using a BCl{sub 3} gas, because the reaction products (e.g., boron oxide) on the reactor walls frequently cause contamination of the product wafers during etching. CH-assisted chlorine cleaning, which is a combination of CHF{sub 3} and Cl{sub 2} plasma treatment, was found to effectively remove Ti residue from the reactor walls. This result shows that CH radicals play an important role in deoxidizing and/or defluorinating Ti residue on the reactor walls.

First-wall-coating candidates for ICF reactor chambers using dry-wall protection only

International Nuclear Information System (INIS)

Sink, D.A.

1983-01-01

Twenty pure metals were considered as potential candidates for first-wall coatings of ICF reactor chambers. Seven were found to merit further consideration based on the results of computer-code calculations of figures-of-merit. The seven are rhenium, iridium, molybdenum, chromium, tungsten, tantalum, and niobium (listed in order of decreasing values of figures-of-merit). The calculations are based on mechanical, thermal, and vacuum vaporization engineering constraints. A number of alloys of these seven metals are suggested as additional candidates

Lifetime analysis for fusion reactor first walls and divertor plates

International Nuclear Information System (INIS)

Horie, T.; Tsujimura, S.; Minato, A.; Tone, T.

1987-01-01

Lifetime analysis of fusion reactor first walls and divertor plates is performed by (1) a one-dimensional analytical plate model, and (2) a two-dimensional elastic-plastic finite element method. Life-limiting mechanisms and the limits of applicability for these analysis methods are examined. Structural design criteria are also discussed. (orig.)

Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

International Nuclear Information System (INIS)

Merola, M.; Biggio, M.

1989-01-01

Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

Influence of natural convection and diluent inerting on H2 and CO oxidation in the reactor cavity

International Nuclear Information System (INIS)

Wong, C.C.

1988-01-01

The question of complete in-cavity oxidation of combustible gases produced by core-concrete interactions following vessel breach has been investigated. It is overly optimistic to assume a complete oxidation because a variety of phenomena, such as steam inerting and oxygen transport by natural convection, may influence the degree of in-cavity oxidation that takes place. HECTR analyses of an ice-condenser containment during an S2HF drain-closed accident show that the in-cavity oxidation process is limited by the rate at which oxygen is transported into the reactor cavity region. Accumulation and subsequent combustion of hydrogen and carbon monoxide in the upper and lower compartments generate a peak pressure of 384 kPa (56 psig) at 7.4 h, that an earlier IDCOR analysis did not predict. (orig.)

Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

International Nuclear Information System (INIS)

Wolf, D.

1987-01-01

The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper

Two-dimensional nucleonics calculations for a ''FIRST STEP'' conceptual ICF reactor

International Nuclear Information System (INIS)

Davidson, J.W.; Battat, M.E.; Saylor, W.W.; Pendergrass, J.H.; Dudziak, D.J.

1985-01-01

A detailed two-dimensional nucleonic analysis has been performed for the FIRST STEP conceptual ICF reactor blanket design. The reactor concept incorporated in this design is a modified wetted-wall cavity with target illumination geometry left as a design variable. The 2-m radius spherical cavity is surrounded by a blanket containing lithium and 238 U as fertile species and also as energy multipliers. The blanket is configured as 0.6-m-thick cylindrical annuli containing modified LMFBR-type fuel elements with 0.5-m-thick fuel-bearing axial end plugs. Liquid lithium surrounds the inner blanket regions and serves as the coolant for both the blanket and the first wall. The two-dimensional analysis of the blanket performance was made using the 2-D discrete-ordinates code TRISM, and benchmarked with the 3-D Monte Carlo code MCNP. Integral responses including the tritium breeding ratio (TBR), plutonium breeding ratio (PUBR), and blanket energy multiplication were calculated for axial and radial blanket regions. Spatial distributions were calculated for steady-state rates of fission, neutron heating, prompt gamma-ray heating, and fuel breeding

Insulated Masonry Cavity Walls. Proceedings of the Research Correlation Conference by the Building Research Institute, Division of Engineering and Industrial Research. (April 1960).

Science.gov (United States)

National Academy of Sciences - National Research Council, Washington, DC.

Publication of conference paper texts include --(1) history and development of masonry cavity walls, (2) recent research related to determination of thermal and moisture resistance, (3) wall design and detailing, (4) design for crack prevention, (5) mortar specification characteristics, (6) performance experience with low-rise buildings, (7)…

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

First wall thermal--mechanical analyses of the reference theta-pinch reactor

International Nuclear Information System (INIS)

Krakowski, R.A.; Hagenson, R.L.; Cort, G.E.

1977-01-01

The thermal-mechanical response of the Reference Theta-Pinch Reactor (RTPR) first wall was analyzed. The first wall problems anticipated for a pulsed, high-β fusion power plant can be ameliorated by either alterations in the physics operating point, materials reengineering, or blanket/first wall reconfiguration. Within the latter ''configuration'' scenario, a two-fold approach has been adopted for the thermal-mechanical portion of the RTPR first wall technology assessment. First, a number of new first wall configurations (bonded or unbonded laminated composites, all-ceramic structures, protective and/or sacrificial ''bumpers'') were considered. Second, a more quantitative failure criterion, based on the developing theories of fracture mechanics, was identified. For each first wall configuration, transient heat transfer and thermoelastic stress calculations have been made. Two-dimensional finite element structural analyses have been made for a variety of mechanical boundary conditions. Only the Al 2 O 3 /Nb - 1 Zr system has been considered. The results of this study indicated a wide range of design solutions to the pulsed thermal stress problem anticipated for the RTPR

Heat and mass transfer in the HYLIFE ICF reactor cavity

International Nuclear Information System (INIS)

Glenn, L.A.

1981-01-01

A quasi-one dimensional method was developed for calculating transient, compressible, viscous flow across a complex array of tubes or jets. The method also accounts for the diffusion of radiation and for heat and mass exchange between the fluid and the jets. The application was to the impulsive crossflow of a lithium plasma through a close-packed annular arrangement of liquid lithium jets, a problem that arises in the design of inertial confinement fusion reactors. It was found that approximately 2/3 of the energy initially contained in the plasma will diffuse into the liquid jets, not including an additional 7-10% which will go towards jet surface vaporization. Nevertheless, the peak hoop stress in the first wall of the reactor appears to derive from direct impact of the plasma, rather than from the subsequent impact of the jets or fragments thereof. (orig.)

BIOREACTOR WITH LID FOR EASY ACCESS TO INCUBATION CAVITY

DEFF Research Database (Denmark)

2012-01-01

There is provided a bioreactor which is provided with a lid (13) that facilitates access to the incubation cavity. Specifically the end wall of the incubation cavity is constituted by the lid (13) so that removal of the cap renders the incubation cavity fully accessible.......There is provided a bioreactor which is provided with a lid (13) that facilitates access to the incubation cavity. Specifically the end wall of the incubation cavity is constituted by the lid (13) so that removal of the cap renders the incubation cavity fully accessible....

Reactor design considerations for inertial confinement fusion

International Nuclear Information System (INIS)

Booth, L.A.

1979-01-01

The most challenging reactor design consideration is protection of the cavity wall from the various energy forms as released by the pellet and as affected by the reaction-chamber phenomena. These phenomena depend on both the design and the yield of the pellet, as well as on ambient conditions in the chamber at the time of the pellet microexplosion. The effects on pellet energy-release mechanisms of various reaction chamber atmosphere options are summarized

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Fuel element for high-temperature nuclear power reactors

International Nuclear Information System (INIS)

Schloesser, J.

1974-01-01

The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

First-wall and blanket engineering development for magnetic-fusion reactors

International Nuclear Information System (INIS)

Baker, C.; Herman, H.; Maroni, V.; Turner, L.; Clemmer, R.; Finn, P.; Johnson, C.; Abdou, M.

1981-01-01

A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories

Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

International Nuclear Information System (INIS)

Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

1996-01-01

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

Steam exit flow design for aft cavities of an airfoil

Science.gov (United States)

Storey, James Michael; Tesh, Stephen William

2002-01-01

Turbine stator vane segments have inner and outer walls with vanes extending therebetween. The inner and outer walls have impingement plates. Steam flowing into the outer wall passes through the impingement plate for impingement cooling of the outer wall surface. The spent impingement steam flows into cavities of the vane having inserts for impingement cooling the walls of the vane. The steam passes into the inner wall and through the impingement plate for impingement cooling of the inner wall surface and for return through return cavities having inserts for impingement cooling of the vane surfaces. A skirt or flange structure is provided for shielding the steam cooling impingement holes adjacent the inner wall aerofoil fillet region of the nozzle from the steam flow exiting the aft nozzle cavities. Moreover, the gap between the flash rib boss and the cavity insert is controlled to minimize the flow of post impingement cooling media therebetween. This substantially confines outflow to that exiting via the return channels, thus furthermore minimizing flow in the vicinity of the aerofoil fillet region that may adversely affect impingement cooling thereof.

Some stress-related issues in tokamak fusion reactor first walls

International Nuclear Information System (INIS)

Majumdar, S.; Pai, B.; Ryder, R.H.

1987-01-01

Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

Water injection device for reactor container

International Nuclear Information System (INIS)

Sakaki, Isao.

1996-01-01

A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

Analysis of AP1000{sup TM} reactor vessel cavity and support cooling

Energy Technology Data Exchange (ETDEWEB)

Craig, K.J. [Westinghouse Electric South Africa, 32 Park Avenue North, Highway Business Park, Centurion, 0157 (SOUTH AFRICA); Harkness, A.W. [Nuclear Power Plants, Westinghouse Electric Company, LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Kritzinger, H.P.; Hoffmann, J.E. [Pebble Bed Modular Reactor (Pty) Ltd, 1279 Mike Crawford Avenue, Centurion (South Africa)

2010-07-01

The paper investigates a Computational Fluid Dynamic (CFD) analysis of the air cooling of the Reactor Vessel (RV) cavity and RV supports. All the Heating, Ventilation and Air Conditioning (HVAC) flow of the RV cavity has to pass through the four RV supports supporting the four cold legs (cold inlets from the two steam generators) of the AP1000{sup TM} reactor. The RV support has a complex flow path leading to significant pressure drops to provide the necessary cooling. The insulation surrounding the RV has a specification on the amount of heat that may be transferred (lost) from the RV in order to maximize the heat transfer to the coolant driving the steam generators. This heat loss is applied as a boundary condition to the solution domain. Another heat source that is considered is that due to nuclear heating. Due to the fact that the heat source is nuclear in nature, gamma and neutron heating have to be considered for the surrounding structures. These include the carbon steel structural module that encapsulates the RV cavity, as well as the concrete poured around this module. The space in the gap between the RV insulation and the structural module steel shell is not only obstructed by the insulation supports, but also by wells or tubes within which power and intermediate ex-core detectors are located. Source-range ex-core detectors are embedded in the concrete surrounding the structural module. All these detectors have a limited operating temperature range, and together with limits on concrete temperatures for safety considerations, necessitate the need for CFD simulations to determine the range of operational temperatures seen by these components. The CFD simulations also provide an estimate of the pressure drop through the cavity between the RV insulation and structural module, as well as that through the four RV supports. Results presented include ANSYS{sup R} FLUENT{sup R} simulations describing the modelling procedure that was followed, namely to combine

Protective interior wall and attaching means for a fusion reactor vacuum vessel

International Nuclear Information System (INIS)

Phelps, R.D.; Upham, G.A.; Anderson, P.M.

1988-01-01

A protective wall for the interior surface of a fusion reactor vessel wall is described comprising: an array of plates, each plate of the array including a main body section, a pair of edge sections bent at an angle with respect to the main body section, and a pair of flange-like end sections each having protruding sections with cut-aways therein, the protruding sections of the flange-like end sections extending in a direction substantially parallel to the main body section; and means operatively associated with the protruding sections of the flange-like end sections of the plates for mounting the array of plates to an associated vessel wall to be protected

SIMS investigations of wall coatings for application in nuclear fusion reactors

International Nuclear Information System (INIS)

Friedbacher, G.; Virag, A.; Grasserbauer, M.; Esser, H.G.; Wienhold, P.

1989-01-01

Carbon coated metals have proven to be useful materials for limiters and the first wall in fusion reactors. In this paper SIMS investigations of a-C:B single coated and a-C:D/a-C:B double coated stainless steel samples, which have been exposed to TOKAMAK discharges in deuterium and helium, are described. (orig.)

Numerical study of the behavior of methane-hydrogen/air pre-mixed flame in a micro reactor equipped with catalytic segmented bluff body

International Nuclear Information System (INIS)

Baigmohammadi, Mohammadreza; Tabejamaat, Sadegh; Zarvandi, Jalal

2015-01-01

In this work, combustion characteristics of premixed methane-hydrogen/air in a micro reactor equipped with a catalytic bluff body is investigated numerically. In this regard, the detailed chemistry schemes for gas phase (homogeneous) and the catalyst surface (heterogeneous) are used. The applied catalytic bluff body is coated with a thin layer of platinum (Pt) on its surface. Also, the lean reactive mixture is entered to the reactor with equivalence ratio 0.9. The results of this study showed that the use of catalytic bluff body in the center of a micro reactor can significantly increase the flame stability, especially at high velocities. Moreover, it is found that a catalytic bluff body with several cavities on its surface and also high thermal conductivity improves the flame stability more than a catalytic bluff body without cavities and low thermal conductivity. Finally, it is maintained that the most advantage of using the catalytic bluff body is its easy manufacturing process as compared to the catalytic wall. This matter seems to be more prevalent when we want to create several cavities with various sizes on the bluff-body. - Highlights: • Presence of a bluff body in a micro reactor can move the flame towards the upstream. • Catalytic bluff body can significantly increase flame stability at high velocities. • Creating non-catalytic cavities on the bluff body promotes homogeneous reactions. • Segmented catalytic bluff body improves the flame stability more than a simple one. • Creating the segments on a bluff body is easier compared to a wall

Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

2016-10-15

A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

Damage of first wall materials in fusion reactors under nonstationary thermal effects

International Nuclear Information System (INIS)

Maslaev, S.A.; Platonov, Yu.M.; Pimenov, V.N.

1991-01-01

The temperature distribution in the first wall of a fusion reactor was calculated for nonstationary thermal effects of the type of plasma destruction or the flow of 'running electrons' taking into account the melting of the surface layer of the material. The thickness of the resultant damaged layer in which thermal stresses were higher than the tensile strength of the material is estimated. The results were obtained for corrosion-resisting steel, aluminium and vanadium. Flowing down of the molten layer of the material of the first wall is calculated. (author)

First wall and blanket design for the STARFIRE commercial tokamak power reactor

International Nuclear Information System (INIS)

Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

1979-01-01

The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

Development of a porous wall reactor for Oxidation in Supercritical Water. Hydrodynamic Modelling and application to salty wastes

International Nuclear Information System (INIS)

Fauvel, E.

2002-01-01

This report deals with a transpiring wall reactor for supercritical water oxidation of organic effluents. The singularity of the reactor lies on the inner porous tube made of alumina to minimise both limiting problems, corrosion and salt precipitation. The presence of the inner tube implies a rather complex hydrodynamics. Thus, an hydrodynamic study was performed, in an original way, in a supercritical fluid using the method of the residence time distribution. It enabled to determine the hydrodynamic model of the reactor. Moreover, an inspecting device of the resistance of the inner tube to thermal gradients was developed. Lastly, the performances of the transpiring wall reactor were tested on model compounds such as sodium sulphate and the mixture of dodecane/tributylphosphate. (author) [fr

An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

International Nuclear Information System (INIS)

Murphy, J.G.; Corradini, M.L.

1997-01-01

The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

Geometric optimization of a solar cubic-cavity multi-tubular thermochemical reactor using a Monte Carlo-finite element radiative transfer model

International Nuclear Information System (INIS)

Valades-Pelayo, P.J.; Romero-Paredes, H.; Arancibia-Bulnes, C.A.; Villafán-Vidales, H.I.

2016-01-01

In the present study, the optimization of a multi-tubular solar thermochemical cavity reactor is carried out. The reactor consists of a cubic cavity made of woven graphite, housing nine 2.54 cm diameter tungsten tubes. A heat transfer model is developed and implemented considering high-temperature radiative transfer at steady state. The temperature distribution on the receiver tubes is determined by using a hybrid Monte Carlo-finite volume approach. The optimization aims at maximizing average tube temperature by varying tube locations. Optimal tube distributions are explored by using a custom-made stochastic, multi-parameter, global optimization algorithm. A considerable increase in average temperature as well as improvement on temperature uniformity is found in the optimized tube arrays. Patterns among the different optimal distributions are found, and general features are discussed.

Modelling of film condensation on the reactor containment walls

International Nuclear Information System (INIS)

Leduc, Christian

1995-01-01

A containment code used in nuclear plant safety analysis must be able to predict evolutions of steam, air and hydrogen concentrations and pressure in the containment of a pressurized water reactor in an accidental situation. Steam condensation on cold walls is an essential factor for these evolutions as it allows the release of an important heat flow, and locally reduces steam concentration. In this research thesis, the author proposes a film condensation model in presence of un-condensable gases. The film flow is supposed to be laminar. Three different approaches are used to model transfers in boundary layers: global correlations in which a hybrid Grashof number is used which expresses the mass and thermal nature of convection, a boundary layer calculation using wall rules for a forced convection regime, and a boundary layer calculation using a k-epsilon model with a low Reynolds number for a natural convection regime. Each approach requires very different mesh fineness at the vicinity of the wall. Models are implemented in the 3-D TRIO-VF thermo-hydraulic code. The obtained theoretical heat transfer coefficients are compared with experimental results [fr

Mixed convection heat transfer enhancement in a cubic lid-driven cavity containing a rotating cylinder through the introduction of artificial roughness on the heated wall

Science.gov (United States)

Kareem, Ali Khaleel; Gao, Shian

2018-02-01

The aim of the present numerical investigation is to comprehensively analyse and understand the heat transfer enhancement process using a roughened, heated bottom wall with two artificial rib types (R-s and R-c) due to unsteady mixed convection heat transfer in a 3D moving top wall enclosure that has a central rotating cylinder, and to compare these cases with the smooth bottom wall case. These different cases (roughened and smooth bottom walls) are considered at various clockwise and anticlockwise rotational speeds, -5 ≤ Ω ≤ 5, and Reynolds numbers of 5000 and 10 000. The top and bottom walls of the lid-driven cavity are differentially heated, whilst the remaining cavity walls are assumed to be stationary and adiabatic. A standard k-ɛ model for the Unsteady Reynolds-Averaged Navier-Stokes equations is used to deal with the turbulent flow. The heat transfer improvement is carefully considered and analysed through the detailed examinations of the flow and thermal fields, the turbulent kinetic energy, the mean velocity profiles, the wall shear stresses, and the local and average Nusselt numbers. It has been concluded that artificial roughness can strongly affect the thermal fields and fluid flow patterns. Ultimately, the heat transfer rate has been dramatically increased by involving the introduced artificial rips. Increasing the cylinder rotational speed or Reynolds number can enhance the heat transfer process, especially when the wall roughness exists.

TEM observations of crack tip: cavity interactions

International Nuclear Information System (INIS)

Horton, J.A.; Ohr, S.M.; Jesser, W.A.

1981-01-01

Crack tip-cavity interactions have been studied by performing room temperature deformation experiments in a transmission electron microscope on ion-irradiated type 316 stainless steel with small helium containing cavities. Slip dislocations emitted from a crack tip cut, sheared, and thereby elongated cavities without a volume enlargement. As the crack tip approached, a cavity volume enlargement occurred. Instead of the cavities continuing to enlarge until they touch, the walls between the cavities fractured. Fracture surface dimples do not correlate in size or density with these enlarged cavities

High Performance Walls in Hot-Dry Climates

Energy Technology Data Exchange (ETDEWEB)

Hoeschele, Marc [National Renewable Energy Lab. (NREL), Golden, CO (United States); Springer, David [National Renewable Energy Lab. (NREL), Golden, CO (United States); Dakin, Bill [National Renewable Energy Lab. (NREL), Golden, CO (United States); German, Alea [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2015-01-01

High performance walls represent a high priority measure for moving the next generation of new homes to the Zero Net Energy performance level. The primary goal in improving wall thermal performance revolves around increasing the wall framing from 2x4 to 2x6, adding more cavity and exterior rigid insulation, achieving insulation installation criteria meeting ENERGY STAR's thermal bypass checklist, and reducing the amount of wood penetrating the wall cavity.

Development of a coupling code for PWR reactor cavity radiation streaming calculation

International Nuclear Information System (INIS)

Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

2012-01-01

PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

Coupling of an overdriven cavity

International Nuclear Information System (INIS)

Garbin, H.D.

1993-01-01

It is well known that when a nuclear test is conducted in a sufficiently large cavity, the resulting seismic signal is sharply reduced when compared to a normal tamped event. Cavity explosions are of interest in the seismic verification community because of this possibility of reducing the seismic energy generated which can lower signal amplitudes and make detection difficult. Reduced amplitudes would also lower seismic yield estimates which has implications in a Threshold Test Ban Treaty (TTBT). In the past several years, there have been a number of nuclear tests at NTS (Nevada Test Site) inside hemispherical cavities. Two such tests were MILL YARD and MISTY ECHO which had instrumentation at the surface and in the free-field. These two tests differ in one important aspect. MILL YARD was completely decoupled i.e., the cavity wall behaved in an elastic manner. It was estimated that MILL YARD's ground motion was reduced by a factor of at least 70. In contrast, MISTY ECHO was detonated in a hemispherical cavity with the same dimensions as MILL YARD, but with a much larger device yield. This caused an inelastic behavior on the wall and the explosion was not fully decoupled

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

Energy Technology Data Exchange (ETDEWEB)

Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

2015-04-29

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

Development of a helical-coil double wall tube steam generator for 4S reactor

International Nuclear Information System (INIS)

Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

2011-01-01

The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

International Nuclear Information System (INIS)

Ogawa, Y; Goto, T; Okano, K; Asaoka, Y; Hiwatari, R; Someya, Y

2008-01-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G∼100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ∼ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

Energy Technology Data Exchange (ETDEWEB)

Ogawa, Y [High Temperature Plasma Center, University of Tokyo, Chiba (Japan); Goto, T; Okano, K [Graduate School of Frontier Sciences, University of Tokyo, Chiba (Japan); Asaoka, Y; Hiwatari, R [Central Research Institute for Electric Power Industry, Komae, Tokyo (Japan); Someya, Y [Graduate School of Engineering, Musashi Institute of Technology, Tokyo (Japan)], E-mail: ogawa@ppl.k.u-tokyo.ac.jp

2008-05-15

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G{approx}100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 {approx} 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

Science.gov (United States)

Ogawa, Y.; Goto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Someya, Y.

2008-05-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Effects of Active and Passive Control Techniques on Mach 1.5 Cavity Flow Dynamics

Directory of Open Access Journals (Sweden)

Selin Aradag

2017-01-01

Full Text Available Supersonic flow over cavities has been of interest since 1960s because cavities represent the bomb bays of aircraft. The flow is transient, turbulent, and complicated. Pressure fluctuations inside the cavity can impede successful weapon release. The objective of this study is to use active and passive control methods on supersonic cavity flow numerically to decrease or eliminate pressure oscillations. Jet blowing at several locations on the front and aft walls of the cavity configuration is used as an active control method. Several techniques are used for passive control including using a cover plate to separate the flow dynamics inside and outside of the cavity, trailing edge wall modifications, such as inclination of the trailing edge, and providing curvature to the trailing edge wall. The results of active and passive control techniques are compared with the baseline case in terms of pressure fluctuations, sound pressure levels at the leading edge, trailing edge walls, and cavity floor and in terms of formation of the flow structures and the results are presented. It is observed from the results that modification of the trailing edge wall is the most effective of the control methods tested leading to up to 40 dB reductions in cavity tones.

Behavior of a heavy cylinder in a horizontal cylindrical liquid-filled cavity at modulated rotation

International Nuclear Information System (INIS)

Kozlov, Nikolai V; Vlasova, Olga A

2016-01-01

The behavior of a heavy cylindrical solid in a horizontal cylindrical cavity is experimentally investigated. The cavity is filled with a viscous liquid and rotates. Two rotation regimes are considered. The first one is steady rotation. A number of body motion regimes are found depending on the cavity rotation speed. The second regime is a modulated rotation, in which the rotation speed is varying periodically. It can be presented as a sum of steady rotation and librations. On the whole, three different cases of the body repulsion from the cavity wall are observed. In the first case, the repulsion occurs when the body slides over a rotating cavity wall. In the second case, the body being in the centrifuged state—when it rotates with the fluid—detaches from the cavity wall under the action of gravity. In the third case, at librations, the wall performs oscillations and the body is repulsed from the wall due to the nonlinear viscous interaction with the fluid. (paper)

Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

Directory of Open Access Journals (Sweden)

Parma Edward J.

2016-01-01

Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

Monte Carlo radiative transfer simulation of a cavity solar reactor for the reduction of cerium oxide

Energy Technology Data Exchange (ETDEWEB)

Villafan-Vidales, H.I.; Arancibia-Bulnes, C.A.; Dehesa-Carrasco, U. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Privada Xochicalco s/n, Col. Centro, A.P. 34, Temixco, Morelos 62580 (Mexico); Romero-Paredes, H. [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco No.186, Col. Vicentina, A.P. 55-534, Mexico D.F 09340 (Mexico)

2009-01-15

Radiative heat transfer in a solar thermochemical reactor for the thermal reduction of cerium oxide is simulated with the Monte Carlo method. The directional characteristics and the power distribution of the concentrated solar radiation that enters the cavity is obtained by carrying out a Monte Carlo ray tracing of a paraboloidal concentrator. It is considered that the reactor contains a gas/particle suspension directly exposed to concentrated solar radiation. The suspension is treated as a non-isothermal, non-gray, absorbing, emitting, and anisotropically scattering medium. The transport coefficients of the particles are obtained from Mie-scattering theory by using the optical properties of cerium oxide. From the simulations, the aperture radius and the particle concentration were optimized to match the characteristics of the considered concentrator. (author)

Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

International Nuclear Information System (INIS)

Oland, C.B.

1985-01-01

Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.

2007-01-01

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

Numerical study of three-dimensional natural convection and entropy generation in a cubical cavity with partially active vertical walls

Directory of Open Access Journals (Sweden)

Abdullah A.A.A Al-Rashed

2017-09-01

Full Text Available Natural convection and entropy generation due to the heat transfer and fluid friction irreversibilities in a three-dimensional cubical cavity with partially heated and cooled vertical walls has been investigated numerically using the finite volume method. Four different arrangements of partially active vertical sidewalls of the cubical cavity are considered. Numerical calculations are carried out for Rayleigh numbers from (103 ≤ Ra ≤ 106, various locations of the partial heating and cooling vertical sidewalls, while the Prandtl number of air is considered constant as Pr=0.7 and the irreversibility coefficient is taken as (φ=10−4. The results explain that the total entropy generation rate increases when the Rayleigh number increases. While, the Bejan number decreases as the Rayleigh number increases. Also, it is found that the arrangements of heating and cooling regions have a significant effect on the fluid flow and heat transfer characteristics of natural convection and entropy generation in a cubical cavity. The Middle-Middle arrangement produces higher values of average Nusselt numbers.

Timber frame walls

DEFF Research Database (Denmark)

Hansen, Ernst Jan de Place; Brandt, Erik

2010-01-01

A ventilated cavity is usually considered good practice for removing moisture behind the cladding of timber framed walls. Timber frame walls with no cavity are a logical alternative as they are slimmer and less expensive to produce and besides the risk of a two-sided fire behind the cladding....... It was found that the specific damages made to the vapour barrier as part of the test did not have any provable effect on the moisture content. In general elements with an intact vapour barrier did not show a critical moisture content at the wind barrier after four years of exposure....

Design and fabrication of foam-insulated cryogenic target for wet-wall laser fusion reactor

International Nuclear Information System (INIS)

Norimatsu, T.; Takeda, T.; Nagai, K.; Mima, K.; Yamanaka, T.

2003-01-01

A foam insulated cryogenic target was proposed for use in a future laser fusion reactor with a wet wall. This scheme can protect the solid DT layer from melting due to surface heating by adsorption of metal vapor without significant reduction in the target gain. Design spaces for the injection velocity and the acceptable vapor pressure in the reactor are discussed. Basic technology to fabricate such structure was demonstrated by emulsion process. Concept of a cryogenic fast-ignition target with a gold guiding cone was proposed together with direct injection filling of liquid DT. (author)

Post-cast EDM method for reducing the thickness of a turbine nozzle wall

Science.gov (United States)

Jones, Raymond Joseph; Bojappa, Parvangada Ganapathy; Kirkpatrick, Francis Lawrence; Schotsch, Margaret Jones; Rajan, Rajiv; Wei, Bin

2002-01-01

A post-cast EDM process is used to remove material from the interior surface of a nozzle vane cavity of a turbine. A thin electrode is passed through the cavity between opposite ends of the nozzle vane and displaced along the interior nozzle wall to remove the material along a predetermined path, thus reducing the thickness of the wall between the cavity and the external surface of the nozzle. In another form, an EDM process employing a profile as an electrode is disposed in the cavity and advanced against the wall to remove material from the wall until the final wall thickness is achieved, with the interior wall surface being complementary to the profile surface.

Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach

International Nuclear Information System (INIS)

Thielman, Jeff; Ge, Ping; Wu, Qiao; Parme, Laurence

2005-01-01

The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units

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.

Solar gasification of biomass: design and characterization of a molten salt gasification reactor

Science.gov (United States)

Hathaway, Brandon Jay

The design and implementation of a prototype molten salt solar reactor for gasification of biomass is a significant milestone in the development of a solar gasification process. The reactor developed in this work allows for 3 kWth operation with an average aperture flux of 1530 suns at salt temperatures of 1200 K with pneumatic injection of ground or powdered dry biomass feedstocks directly into the salt melt. Laboratory scale experiments in an electrically heated reactor demonstrate the benefits of molten salt and the data was evaluated to determine the kinetics of pyrolysis and gasification of biomass or carbon in molten salt. In the presence of molten salt overall gas yields are increased by up to 22%; pyrolysis rates double due to improved heat transfer, while carbon gasification rates increase by an order of magnitude. Existing kinetic models for cellulose pyrolysis fit the data well, while carbon gasification in molten salt follows kinetics modeled with a 2/3 order shrinking-grain model with a pre-exponential factor of 1.5*106 min-1 and activation energy of 158 kJ/mol. A reactor concept is developed based around a concentric cylinder geometry with a cavity-style solar receiver immersed within a volume of molten carbonate salt. Concentrated radiation delivered to the cavity is absorbed in the cavity walls and transferred via convection to the salt volume. Feedstock is delivered into the molten salt volume where biomass gasification reactions will be carried out producing the desired product gas. The features of the cavity receiver/reactor concept are optimized based on modeling of the key physical processes. The cavity absorber geometry is optimized according to a parametric survey of radiative exchange using a Monte Carlo ray tracing model, resulting in a cavity design that achieves absorption efficiencies of 80%-90%. A parametric survey coupling the radiative exchange simulations to a CFD model of molten salt natural convection is used to size the annulus

Laboratory testing of joints between windows and highly insulated cavity walls. Investigations of tightness against rain and wind

Energy Technology Data Exchange (ETDEWEB)

Kjaer, A

1983-10-01

In the Danish energy research programme, 1EFP 80, a number of laboratory tests have been carried out on models of highly insulated cavity brick walls in order to study rain- and wind tightness of the joints between windows and such walls. Tests have been carried out with joints tightened only with a rain barrier as well as with joints according to the two stage joint principle. In the exterior part of the joint has in both cases been used a mortar, and expanding gasket, an EPDM-profile and wooden battens. Further an experiment has been carried out on a plastic window, where mastic was used as well in the exterior as the interior part of the joint. The findings were that a two-stage joint gives the best performance as well regarding air tightness as rain tightness. Further the experiments have shown that a window frame should have a depth of at least 90 mm in order to design a joint between window and wall, which is satisfactory as well regarding thermal insulation as resistance to rain and wind.

Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1977-01-01

A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1978-01-01

A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

EFFECT OF DISCRETE HEATER AT THE VERTICAL WALL OF THE CAVITY OVER THE HEAT TRANSFER AND ENTROPY GENERATION USING LBM

Directory of Open Access Journals (Sweden)

Mousa Farhadi

2011-01-01

Full Text Available In this paper Lattice Boltzmann Method (LBM was employed for investigation the effect of the heater location on flow pattern, heat transfer and entropy generation in a cavity. A 2D thermal lattice Boltzmann model with 9 velocities, D2Q9, is used to solve the thermal flow problem. The simulations were performed for Rayleigh numbers from 103 to 106 at Pr = 0.71. The study was carried out for heater length of 0.4 side wall length which is located at the right side wall. Results are presented in the form of streamlines, temperature contours, Nusselt number and entropy generation curves. Results show that the location of heater has a great effect on the flow pattern and temperature fields in the enclosure and subsequently on entropy generation. The dimensionless entropy generation decreases at high Rayleigh number for all heater positions. The ratio of averaged Nusselt number and dimensionless entropy generation for heater located on vertical and horizontal walls was calculated. Results show that higher heat transfer was observed from the cold walls when the heater located on vertical wall. On the other hand, heat transfer increases from the heater surface when it located on the horizontal wall.

Mechanical and microstructural characterization of low activation steels as first wall of nuclear fusion reactors

International Nuclear Information System (INIS)

Hernandez, M.T.; Lapena, J.; Diego, G. de; Schirra, M.

1996-01-01

Currently, the design development of fusion reactors and the possible materials to use in them are being studied in parallel. One of the most critical problems in this research is the structural materials selection for the first wall and blanket. The aim of the present work is to study three low activation alloys designed in Germany in which niobium has been substituted by tantalum or cerium. The mechanical results show that the alloys containing cerium are in the same order of the low activation materials known to date, but the tantalum doped alloy produces TaC 3 precipitation that destabilizes the matrix and provokes large microstructural changes. This causes a decrease of the mechanical properties at about 600 degree centigree. This fact makes this alloy insuitable for the first wall on fusion reactors, because the working temperature is near 550 degree centigree. (Author) 11 refs

Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

International Nuclear Information System (INIS)

Lafitte, R.; Marchand, J.D.

1979-01-01

The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

Numerical Studies on Natural Convection Heat Losses from Open Cubical Cavities

Directory of Open Access Journals (Sweden)

M. Prakash

2013-01-01

Full Text Available The natural convection heat losses occurring from cubical open cavities are analysed in this paper. Open cubical cavities of sides 0.1 m, 0.2 m, 0.25 m, 0.5 m, and 1 m with constant temperature back wall boundary conditions and opening ratio of 1 are studied. The Fluent CFD software is used to analyse the three-dimensional (3D cavity models. The studies are carried out for cavities with back wall temperatures between 35°C and 100°C. The effect of cavity inclination on the convective loss is analysed for angles of 0° (cavity facing sideways, 30°, 45°, 60°, and 90° (cavity facing vertically downwards. The Rayleigh numbers involved in this study range between 4.5 × 105 and 1.5 × 109. The natural convection loss is found to increase with an increase in back wall temperature. The natural convection loss is observed to decrease with an increase in cavity inclination; the highest convective loss being at 0° and the lowest at 90° inclination. This is observed for all cavities analysed here. Nusselt number correlations involving the effect of Rayleigh number and the cavity inclination angle have been developed from the current studies. These correlations can be used for engineering applications such as electronic cooling, low- and medium-temperature solar thermal systems, passive architecture, and also refrigeration systems.

Diagnostic techniques for measuring temperature transients and stress transients in the first wall of an ICF reactor

International Nuclear Information System (INIS)

Melamed, N.T.; Taylor, L.H.

1983-01-01

The primary challenge in the design of an Inertial Confinement Fusion (ICF) power reactor is to make the first wall survive the frequent explosions of the pellets. Westinghouse has proposed a dry wall design consisting of steel tubes coated with tantalum. This report describes the design of a test chamber and two diagnostic procedures for experimentally determining the reliability of the Westinghouse design. The test chamber simulates the x-ray and ion pulse irradiation of the wall due to a pellet explosion. The diagnostics consist of remote temperature sensing and surface deformation measurements. The chamber and diagnostics can also be used to test other first-wall designs

Present status of inertial confinement fusion reactor design

International Nuclear Information System (INIS)

Mima, Kunioki; Ido, Shunji; Nakai, Sadao.

1986-01-01

Since inertial nuclear fusion reactors do not require high vacuum and high magnetic field, the structure of the reactor cavity becomes markedly simple as compared with tokamak type fusion reactors. In particular, since high vacuum is not necessary, liquid metals such as lithium and lead can be used for the first wall, and the damage of reactor structures by neutrons can be prevented. As for the core, the energy efficiency of lasers is not very high, accordingly it must be designed so that the pellet gain due to nuclear fusion becomes sufficiently high, and typically, the gain coefficient from 100 to 200 is necessary. In this paper, the perspective of pellet gain, the plan from the present status to the practical reactors, and the conceptual design of the practical reactors are discussed. The plan of fuel ignition, energy break-even and high gain by the implosion mode, of which the uncertain factor due to uneven irradiation and instability was limited to the minimum, was clarified. The scenario of the development of laser nuclear fusion reactors is presented, and the concept of the reactor system is shown. The various types of nuclear fusion-fission hybrid reactors are explained. As for the design of inertial fusion power reactors, the engineering characteristics of the core, the conceptual design, water fall type reactors and DD fuel reactors are discussed. (Kako, I.)

Water entry without surface seal: Extended cavity formation

KAUST Repository

Mansoor, Mohammad M.

2014-03-01

We report results from an experimental study of cavity formation during the impact of superhydrophobic spheres onto water. Using a simple splash-guard mechanism, we block the spray emerging during initial contact from closing thus eliminating the phenomenon known as \\'surface seal\\', which typically occurs at Froude numbers Fr= V0 2/(gR0) = O(100). As such, we are able to observe the evolution of a smooth cavity in a more extended parameter space than has been achieved in previous studies. Furthermore, by systematically varying the tank size and sphere diameter, we examine the influence of increasing wall effects on these guarded impact cavities and note the formation of surface undulations with wavelength λ =O(10)cm and acoustic waves λa=O(D0) along the cavity interface, which produce multiple pinch-off points. Acoustic waves are initiated by pressure perturbations, which themselves are generated by the primary cavity pinch-off. Using high-speed particle image velocimetry (PIV) techniques we study the bulk fluid flow for the most constrained geometry and show the larger undulations ( λ =O (10cm)) have a fixed nature with respect to the lab frame. We show that previously deduced scalings for the normalized (primary) pinch-off location (ratio of pinch-off depth to sphere depth at pinch-off time), Hp/H = 1/2, and pinch-off time, τ α (R0/g) 1/2, do not hold for these extended cavities in the presence of strong wall effects (sphere-to-tank diameter ratio), ε = D 0/Dtank 1/16. Instead, we find multiple distinct regimes for values of Hp/H as the observed undulations are induced above the first pinch-off point as the impact speed increases. We also report observations of \\'kinked\\' pinch-off points and the suppression of downward facing jets in the presence of wall effects. Surprisingly, upward facing jets emanating from first cavity pinch-off points evolve into a \\'flat\\' structure at high impact speeds, both in the presence and absence of wall effects. �

Ideal quantum gas in an expanding cavity: nature of nonadiabatic force.

Science.gov (United States)

Nakamura, K; Avazbaev, S K; Sobirov, Z A; Matrasulov, D U; Monnai, T

2011-04-01

We consider a quantum gas of noninteracting particles confined in the expanding cavity and investigate the nature of the nonadiabatic force which is generated from the gas and acts on the cavity wall. First, with use of the time-dependent canonical transformation, which transforms the expanding cavity to the nonexpanding one, we can define the force operator. Second, applying the perturbative theory, which works when the cavity wall begins to move at time origin, we find that the nonadiabatic force is quadratic in the wall velocity and thereby does not break the time-reversal symmetry, in contrast with general belief. Finally, using an assembly of the transitionless quantum states, we obtain the nonadiabatic force exactly. The exact result justifies the validity of both the definition of the force operator and the issue of the perturbative theory. The mysterious mechanism of nonadiabatic transition with the use of transitionless quantum states is also explained. The study is done for both cases of the hard- and soft-wall confinement with the time-dependent confining length. ©2011 American Physical Society

Electrochemical aspects on corrosion in Swedish reactor containments; Elektrokemiska aspekter paa korrosion i svenska reaktorinneslutningar

Energy Technology Data Exchange (ETDEWEB)

Ullberg, Mats [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

2006-10-15

Post-stressed concrete is used in all Swedish nuclear reactor containments. Steel in concrete is normally protected from corrosion by the highly alkaline pore solution in concrete. A passive film develops on the surface of steel in contact with the pore solution. However, corrosion may still occur under special circumstances. It is therefore desirable to monitor the corrosion status of the containment. A review of the corrosion experience with steel in concrete strongly suggests that the potential problem of most concern for the Swedish reactor containments is cavity formation during grouting of tendons and of penetrations in the containment wall. Cavities break the contact between alkaline grout and steel. Corrosion is then possible, provided the relative humidity is high enough. Normal methods for inspection of the corrosion status of steel reinforcement in concrete are not applicable to very heavy structures like reactor containments. Since inspections are difficult to carry out, it is important that they be focused on the most susceptible portions of the containment. This report is an attempt to assemble potentially useful background information. The original intention was to focus on electrochemical methods of investigation. When it was realized that the potential use of electrochemical methods was limited, the scope of the review was broadened. The present as well as previous investigations indicate that nondestructive testing of grouted tendons is the outstanding problem in the condition assessment of Swedish nuclear reactor containments. Grouted tendons are also used in a very large number of bridges built since the early 1950s. The experience gained in connection with bridges has therefore been investigated. The need for a testing method for grouted tendons in bridges has long been strongly felt and development work has been in progress since the early 1970-ies, for example within the Strategic Highway Research Project in the Unite States. Potential

Development of a computer code, PZRTR, for the thermal hydraulic analysis of a multi-cavity cold gas pressurizer for an integral reactor, SMART-P

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Yoon, J

2003-12-01

The concept of a Multi-cavity Cold Gas PressuriZeR (MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 100 .deg. C, which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR, for the Reactor Coolant System (RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators (OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the

Influence of radiation on double conjugate diffusion in a porous cavity

International Nuclear Information System (INIS)

Azeem,; Idris, Mohd Yamani Idna; Khan, T. M. Yunus; Badruddin, Irfan Anjum; Nik-Ghazali, N.

2016-01-01

The current work highlights the effect of radiation on the conjugate heat and mass transfer in a square porous cavity having a solid wall. The solid wall is placed at the center of cavity. The left surface of cavity is maintained at higher temperature T_w and concentration C_w whereas the right surface is maintained at T_c and C_c such that T_w>T_c and Cw>Cc. The top and bottom surfaces are adiabatic. The governing equations are solved with the help of finite element method by making use of triangular elements. The results are discussed with respect to two different heights of solid wall inside the porous medium along with the radiation parameter.

Surface tension effects on the behavior of a cavity growing, collapsing, and rebounding near a rigid wall.

Science.gov (United States)

Zhang, Zhen-yu; Zhang, Hui-sheng

2004-11-01

Surface tension effects on the behavior of a pure vapor cavity or a cavity containing some noncondensible contents, which is growing, collapsing, and rebounding axisymmetrically near a rigid wall, are investigated numerically by the boundary integral method for different values of dimensionless stand-off parameter gamma, buoyancy parameter delta, and surface tension parameter beta. It is found that at the late stage of the collapse, if the resultant action of the Bjerknes force and the buoyancy force is not small, surface tension will not have significant effects on bubble behavior except that the bubble collapse time is shortened and the liquid jet becomes wider. If the resultant action of the two force is small enough, surface tension will have significant and in some cases substantial effects on bubble behavior, such as changing the direction of the liquid jet, making a new liquid jet appear, in some cases preventing the bubble from rebound before jet impact, and in other cases causing the bubble to rebound or even recollapse before jet impact. The mechanism of surface tension effects on the collapsing behavior of a cavity has been analyzed. The mechanisms of some complicated phenomena induced by surface tension effects are illustrated by analysis of the computed velocity fields and pressure contours of the liquid flow outside the bubble at different stages of the bubble evolution.

The pore of the leaf cavity of Azolla species: teat cell differentiation and cell wall projections.

Science.gov (United States)

Veys, P; Lejeune, A; Van Hove, C

2002-02-01

The differentiation of the specialized secretory teat cells of the leaf cavity pore of Azolla species was investigated at the ultrastructural level with emphasis on their peculiar cell wall projections. The results indicated that the projections are formed as soon as the teat cells complete their differentiation and that their production is principally associated with changes in endoplasmic reticulum profiles. The number of projections increases with the teat cell age and is stimulated under salt and P deficiency stresses. Salt stress also promotes their emergence on Azolla species that under normal conditions do not produce projections. Cytochemical tests on different Azolla species showed that the projection composition is almost identical: proteins, acidic polysaccharides, and pectin are always detected. This study revealed that Azolla teat cell projections differ fundamentally from other types of hitherto described cell wall projections that are considered as remnant structures from cell separation. In contrast, in Azolla teat cells projections are actively produced and compounds are excreted by an exocytotic mechanism. The possible role of the projections in the symbiosis of Azolla spp. with Anabaena azollae is discussed.

Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

1994-10-01

As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

Scram device for atomic reactors

International Nuclear Information System (INIS)

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

1978-01-01

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 release 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. In an alternate embodiment, the actuating bellows is connected to the liquid cavity rather than to the gas cavity. (Auth.)

Heat dissipating nuclear reactor

Science.gov (United States)

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

International Nuclear Information System (INIS)

Fukuzawa, Y.

1979-07-01

Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

Tool for cutting locking cups from guide tube mounting screws in a nuclear reactor

International Nuclear Information System (INIS)

Nee, J.D.; Hahn, J.J.

1987-01-01

This patent describes an apparatus for freeing a socket-head screw from a locking cup therefor in a reactor cavity, wherein the locking cup includes a fixed cylindrical side wall encircling the side surface of the screw head and an annular end wall overlying the outer end surface of the screw head. The apparatus consists of: frame means, cylindrical cutter means having a longitudinal axis and having a frustoconical cutting surface with an inner diameter less than the inner diameter of the locking cup side wall and with an outer diameter greater than the outer diameter of the locking cup side wall, and drive means carried by the frame means and coupled to the cutter means for effecting rotation thereof about the axis, the rotating cutter means are operable for severing the locking cup end wall from the locking cup side wall at the junction therebetween when the cutter means is moved against the locking cup substantially coaxially therewith

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

International Nuclear Information System (INIS)

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

Technical tasks in superconducting cavities

Energy Technology Data Exchange (ETDEWEB)

Saito, Kenji [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan)

1997-11-01

The feature of superconducting rf cavities is an extremely small surface resistance on the wall. It brings a large energy saving in the operation, even those are cooled with liquid helium. That also makes possible to operate themselves in a higher field gradient comparing to normal conducting cavities, and brings to make accelerators compact. These merits are very important for the future accelerator engineering which is planed at JAERI for the neutron material science and nuclear waste transmutation. This machine is a high intensity proton linac and uses sc cavities in the medium and high {beta} sections. In this paper, starting R and D of proton superconducting cavities, several important technical points which come from the small surface resistance of sc cavities, are present to succeed it and also differences between the medium and high - {beta} structures are discussed. (author)

Influence of radiation on double conjugate diffusion in a porous cavity

Energy Technology Data Exchange (ETDEWEB)

Azeem,; Idris, Mohd Yamani Idna [Dept. of Computer System & Technology, University of Malaya, Kuala Lumpur (Malaysia); Khan, T. M. Yunus, E-mail: yunus.tatagar@gmail.com [Dept. of Mechanical Engineering, University of Malaya, Kuala Lumpur, 50603 (Malaysia); Dept. of Mechanical Engineering, BVB College of Engineering & Technology, Hubli (India); Badruddin, Irfan Anjum, E-mail: irfan-magami@Rediffmail.com; Nik-Ghazali, N. [Dept. of Mechanical Engineering, University of Malaya, Kuala Lumpur, 50603 (Malaysia)

2016-05-06

The current work highlights the effect of radiation on the conjugate heat and mass transfer in a square porous cavity having a solid wall. The solid wall is placed at the center of cavity. The left surface of cavity is maintained at higher temperature T{sub w} and concentration C{sub w} whereas the right surface is maintained at T{sub c} and C{sub c} such that T{sub w}>T{sub c} and Cw>Cc. The top and bottom surfaces are adiabatic. The governing equations are solved with the help of finite element method by making use of triangular elements. The results are discussed with respect to two different heights of solid wall inside the porous medium along with the radiation parameter.

Actions to reduce radioactive emissions: prevention of containment failure by flooding Containment and Reactor Cavity; Acciones para la reduccion de emisiones radiactivas: prevencion del fallo de la Contencion mediante la inundacion de la Contencion y de la Cavidad del Reactor

Energy Technology Data Exchange (ETDEWEB)

Fornos Herrando, J.

2013-07-01

The reactor cavity of Asco and Vandellos II is dry type, thus a severe accident leading to vessel failure might potentially end up resulting in the loss of containment integrity, depending on the viability to cool the molten core. Therefore, significant radioactive emissions could be released to outside. In the framework of Fukushima Stress Tests, ANAV has analyzed the convenience of carrying out different actions to prevent failure of the containment integrity in order to reduce radioactive emissions. The aim of this paper is to present and describe the main phenomenological aspects associated with two of these actions: containment flooding and reactor cavity flooding.

Reactor advantages of the belt pinch and liquid metal walls

International Nuclear Information System (INIS)

Kotschenreuther, M.; Manickam, J.; Menard, J.; Rappaport, H.; Zheng Linjin; Dorland, B.; Miller, R.; Turnbull, A.

2001-01-01

MHD stability of highly elongated tokamaks (termed a belt pinch) are considered for high bootstrap fraction cases. By employing high triangularity or indentation, and invoking wall stabilization, and β can be increased by a factor of roughly 3 by increasing κ from 2 to 4. Axisymmetric stability up to κ=4 tolerable by employing a shell which conforms more closely to the boundary than in present experiments. Engineering difficulties with a close fitting shell in a reactor environment may be overcome by employing a liquid lithium alloy shell. Rapid metal flows can lead to potentially deleterious plasma shifts and damping of the flow. (author)

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

Science.gov (United States)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Reactor container

International Nuclear Information System (INIS)

Shibata, Satoru; Kawashima, Hiroaki

1984-01-01

Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

Turbine airfoil having near-wall cooling insert

Science.gov (United States)

Martin, Jr., Nicholas F.; Wiebe, David J.

2017-09-12

A turbine airfoil is provided with at least one insert positioned in a cavity in an airfoil interior. The insert extends along a span-wise extent of the turbine airfoil and includes first and second opposite faces. A first near-wall cooling channel is defined between the first face and a pressure sidewall of an airfoil outer wall. A second near-wall cooling channel is defined between the second face and a suction sidewall of the airfoil outer wall. The insert is configured to occupy an inactive volume in the airfoil interior so as to displace a coolant flow in the cavity toward the first and second near-wall cooling channels. A locating feature engages the insert with the outer wall for supporting the insert in position. The locating feature is configured to control flow of the coolant through the first or second near-wall cooling channel.

Stagnation of ablated metal vapor in laser fusion reactor with liquid wall

International Nuclear Information System (INIS)

Norimatsu, T.; Nagatomo, H.; Azechi, H.; Furukawa, H.; Shimada, Y.; Kurahashi, S.; Kunugi, T.; Kajimura, Y.

2010-11-01

In this paper, formation of clusters by ablated materials and those stagnation at the center of a laser fusion reactor with liquid wall are discussed using improved simulation code DECORE. We will report 1) numerical simulation on formation of clusters immediately before the stagnation, 2) preliminary results on the cluster formation at the first bounce of the stagnation, 3) experimental result on the diameter measurement of micro droplets formed in a simulation experiment with back-side irradiation of laser. (author)

RF cavity for the Novosibirsk race-track microtron-recuperator

International Nuclear Information System (INIS)

Gavrilov, N.; Kuptsov, I.; Kurkin, G.; Mironenko, L.; Petrov, V.; Sedlyarov, I.; Veshcherevich, V.

1994-01-01

Geometry, engineering design and characteristics of a 181 MHz RF cavity are described. The cavity has copper clad stainless steel walls and has a Q of 42,000 and a shunt impedance of 8.5 MOhm. The cavities of that type are parts of an RF system of a CW race-track microtron-recuperator (RTMR). 10 refs.; 16 figs.; 1 tab

A foldable electrode array for 3D recording of deep-seated abnormal brain cavities

Science.gov (United States)

Kil, Dries; De Vloo, Philippe; Fierens, Guy; Ceyssens, Frederik; Hunyadi, Borbála; Bertrand, Alexander; Nuttin, Bart; Puers, Robert

2018-06-01

Objective. This study describes the design and microfabrication of a foldable thin-film neural implant and investigates its suitability for electrical recording of deep-lying brain cavity walls. Approach. A new type of foldable neural electrode array is presented, which can be inserted through a cannula. The microfabricated electrode is specifically designed for electrical recording of the cavity wall of thalamic lesions resulting from stroke. The proof-of-concept is demonstrated by measurements in rat brain cavities. On implantation, the electrode array unfolds in the brain cavity, contacting the cavity walls and allowing recording at multiple anatomical locations. A three-layer microfabrication process based on UV-lithography and Reactive Ion Etching is described. Electrochemical characterization of the electrode is performed in addition to an in vivo experiment in which the implantation procedure and the unfolding of the electrode are tested and visualized. Main results. Electrochemical characterization validated the suitability of the electrode for in vivo use. CT imaging confirmed the unfolding of the electrode in the brain cavity and analysis of recorded local field potentials showed the ability to record neural signals of biological origin. Significance. The conducted research confirms that it is possible to record neural activity from the inside wall of brain cavities at various anatomical locations after a single implantation procedure. This opens up possibilities towards research of abnormal brain cavities and the clinical conditions associated with them, such as central post-stroke pain.

In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

International Nuclear Information System (INIS)

Todt, W.H. Sr.; Playfoot, K.C.

1988-01-01

This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

Conduction and convection heat transfer characteristics of water-based au nanofluids in a square cavity with differentially heated side walls subjected to constant temperatures

Directory of Open Access Journals (Sweden)

Ternik Primož

2014-01-01

Full Text Available The present work deals with the natural convection in a square cavity filled with the water-based Au nanofluid. The cavity is heated on the vertical and cooled from the adjacent wall, while the other two horizontal walls are adiabatic. The governing differential equations have been solved by the standard finite volume method and the hydrodynamic and thermal fields were coupled together using the Boussinesq approximation. The main objective of this study is to investigate the influence of the nanoparticles’ volume fraction on the heat transfer characteristics of Au nanofluids at the given base fluid’s (i.e. water Rayleigh number. Accurate results are presented over a wide range of the base fluid Rayleigh number and the volume fraction of Au nanoparticles. It is shown that adding nanoparticles in a base fluid delays the onset of convection. Contrary to what is argued by many authors, we show by numerical simulations that the use of nanofluids can reduce the heat transfer rate instead of increasing it.

Remote through-wall sampling of the Trawsfynydd reactor pressure vessel: an overview

International Nuclear Information System (INIS)

Curry, A.; Clayton, R.

1996-01-01

This paper summarises the application of robotic equipment for gaining access to and removing through-wall samples from welds of the reactor pressure vessel at Trawsfynydd power station. The environment, which presents hazards due to ionising radiation, radioactive contamination and asbestos bearing materials is described. The means of access, by use of remote vehicles complete with robotic manipulators supported by additional vehicles, is reviewed. The use of Abrasive Water Jet Cutting for sample removal is introduced. The relative advantages and disadvantages of this technique are discussed. (Author)

Movement of the lacrimal canalicular wall under intracanalicular pressure changes observed with dacryoendoscopy.

Science.gov (United States)

Kakizaki, Hirohiko; Takahashi, Yasuhiro; Mito, Hidenori; Nakamura, Yasuhisa

2015-01-01

Movement of the lacrimal canalicular wall has been speculated to occur during blinking. Movement of the common internal ostium has been observed under nasal endoscopy, and pressure changes in the lacrimal canalicular cavity have been observed with a pressure sensor; however, lacrimal canalicular wall movement under pressure changes has not been observed. To examine movement of the lacrimal canalicular wall under intracanalicular pressure changes using dacryoendoscopy. The authors examined 20 obstruction-free lacrimal canaliculi in 10 patients. A dacryoendoscope was inserted, and water was poured into the intracanalicular cavity via the dacryoendoscope's water channel. The water was then poured or suctioned to cause positive or negative pressure changes in the intracanalicular cavity, and movement of the lacrimal canalicular wall was examined. The lacrimal canalicular wall moved flexibly with pressure changes. Under positive pressure, the intracanalicular cavity was dilated; however, it narrowed under negative pressure. The extent of movement was more dramatic in the common canalicular portion than the proximal canalicular portion. Intracanalicular pressure changes cause movement of the lacrimal canalicular wall. There was a consistent relationship between intracanalicular cavity changes and pressure changes, possibly contributing to lacrimal drainage of the canaliculus.

Development of a Computer Code, PZRTR rev 1, for the Thermal Hydraulic Analysis of a Multi-Cavity Cold Gas Pressurizer for an Integral Reactor, SMART-P

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Kang, H. O.; Yoon, J.; Kim, K. K

2006-12-15

The concept of a Multi-cavity Cold Gas PressuriZeR(MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 120 .deg. C which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR rev 1, for the Reactor Coolant System(RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators(OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR rev 1 code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the

Investigation of cascade-type falling liquid-film along first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, T.; Nakai, T.; Kawara, Z.; Norimatsu, T.; Kozaki, Y.

2008-01-01

To protect the first wall of an inertia fusion reactor from extremely high heat flux, X-rays, alpha particles and fuel debris caused by a nuclear fusion reaction, a 'cascade-type' falling liquid-film flow is proposed as a 'liquid-wall' concept. The flow visualization experiment to investigate the feasibility of this liquid-wall concept has been conducted. The preliminary numerical simulation results suggest that the current cascade structure design should be improved because less thermal-mixing is expected. The cascade-type structure has, therefore, been redesigned. This new cascade-type first wall consists of a liquid reservoir which has a free-surface to maintain a constant water head in the rear, and connects to a slit composed of two plates, i.e., the first wall is connected to a slit which is partially made up of the first wall to begin with it. The numerical simulations were performed on the new cascade-type first wall and they show the stable liquid-film flow on it. Moreover, the POP (proof-of-principle) flow visualization experiments, which satisfy the Weber number coincident condition, are carried out using water as the working fluid. By comparing the numerical and experimental results, it was found that the liquid-film flow with 3-5 mm thickness could be stably established. According to these results for the new cascade-type first wall concept, it was confirmed that the coolant flow rate and the thickness of the liquid-film could be controlled if the Weber number coincident condition was satisfied

Natural convection in a cubical cavity with a coaxial heated cylinder

Energy Technology Data Exchange (ETDEWEB)

Aithal, S. M.

2018-03-01

High-resolution three-dimensional simulations were conducted to investigate the velocity and temperature fields in a cold cubical cavity due to natural convection induced by a centrally placed hot cylinder. Unsteady, incompressible Navier-Stokes equations were solved by using a spectral- element method for Rayleigh numbers ranging from 103 to 109. The effect of spanwise thermal boundary conditions, aspect ratio (radius of the cylinder to the side of the cavity), and spanwise temperature distribution of the inner cylinder on the velocity and thermal fields were investigated for each Rayleigh number. Results from two-dimensional calculations were compared with three-dimensional simulations. The 3D results indicate a complex flow structure in the vicinity of the spanwise walls. The results also show that the imposed thermal wall boundary condition impacts the flow and temperature fields strongly near the spanwise walls. The variation of the local Nusselt number on the cylinder surface and enclosure walls at various spanwise locations was also investigated. The local Nusselt number on the cylinder surface and enclosure walls at the cavity mid-plane (Z = 0) is close to 2D simulations for 103 ≤ Ra ≤ 108. Simulations also show a variation in the local Nusselt number, on both the cylinder surface and the enclosure walls, in the spanwise direction, for all Rayleigh numbers studied in this work. The results also indicate that if the enclosure walls are insulated in the spanwise direction (as opposed to a constant temperature), the peak Nusselt number on the enclosure surface occurs near the spanwise walls and is about 20% higher than the peak Nusselt number at the cavity mid-plane. The temporal characteristics of 3D flows are also different from 2D results for Ra > 108. These results suggest that 3D simulations would be more appropriate for flows with Ra > 108.

Investigation on flow oscillation modes and aero-acoustics generation mechanism in cavity

Science.gov (United States)

Yang, Dang-Guo; Lu, Bo; Cai, Jin-Sheng; Wu, Jun-Qiang; Qu, Kun; Liu, Jun

2018-05-01

Unsteady flow and multi-scale vortex transformation inside a cavity of L/D = 6 (ratio of length to depth) at Ma = 0.9 and 1.5 were studied using the numerical simulation method of modified delayed detached eddy simulation (DDES) in this paper. Aero-acoustic characteristics for the cavity at same flow conditions were obtained by the numerical method and 0.6 m by 0.6 m transonic and supersonic wind-tunnel experiments. The analysis on the computational and experimental results indicates that some vortex generates from flow separation in shear-layer over the cavity, and the vortex moves from forward to downward of the cavity at some velocity, and impingement of the vortex and the rear-wall of the cavity occurs. Some sound waves spread abroad to the cavity fore-wall, which induces some new vortex generation, and the vortex sheds, moves and impinges on the cavity rear-wall. New sound waves occur. The research results indicate that sound wave feedback created by the impingement of the shedding-vortices and rear cavity face leads to flow oscillations and noise generation inside the cavity. Analysis on aero-acoustic characteristics inside the cavity is feasible. The simulated self-sustained flow-oscillation modes and peak sound pressure on typical frequencies inside the cavity agree well with Rossiter’s and Heller’s predicated results. Moreover, the peak sound pressure occurs in the first and second flow-oscillation modes and most of sound energy focuses on the low-frequency region. Compared with subsonic speed (Ma = 0.9), aerodynamic noise is more intense at Ma = 1.5, which is induced by compression wave or shock wave in near region of fore and rear cavity face.

Transient mixed convection in a channel with an open cavity filled with porous media

International Nuclear Information System (INIS)

Buonomo, B; Cresci, G; Manca, O; Mesolella, P; Nardini, S

2012-01-01

In this work transient mixed convection in a porous medium in a horizontal channel with a open cavity below is studied numerically. The cavity presents a heated wall at uniform heat flux and the other walls of the cavity and the channel are assumed adiabatic. Air flows through the horizontal channel. The heated wall of the cavity experiences a uniform heat flux in such a way that the forced flow is perpendicular to the motion due to natural convection. The study is carried out employing Brinkman-Forchheimer-extended Darcy model and two energy equations due to the local thermal non-equilibrium assumption. The flow in the channel is assumed to be two-dimensional, laminar, incompressible. Boussinesq approximation is considered. The thermophysical properties of the fluid are evaluated at the ambient temperature. The results for stream function and temperature distribution given at different times are obtained. Wall temperature value are given and also, the velocity and temperature profiles in several sections of the cavity are presented. In addition, the Nusselt number, both local and average, is presented along with the temporal variations of the average Nusselt number.

An introduction to the analysis of multi-cavity prestressed concrete pressure vessels

International Nuclear Information System (INIS)

Silva, M.C.A.T. da.

1986-01-01

The present work is a study of multi-cavity prestressed concrete pressure vessels (PCRV) for nuclear reactors. A review is made of the designs, analises and models of multi-cavity concrete pressure vessels. A preliminary evaluation of the NONSAP program for applications in complex three-dimensional structures such as a multi-cavity pressure vessel is also made. A model of a PCRV of a 1000 MW(e) high-temperature gas cooled reactor was selected for a three-dimensional analysis with the NONSAP program. The results obtained are compared with experimental data. (Author) [pt

Maximum attainable power density and wall load in tokamaks underlying reactor relevant constraints

International Nuclear Information System (INIS)

Borrass, K.; Buende, R.

1979-09-01

The characteristic data of tokamaks optimized with respect to their power density or wall load are determined. Reactor relevant constraints are imposed, such as a fixed plant net power output, a fixed blanket thickness and the dependence of the maximum toroidal field on the geometry and conductor material. The impact of finite burn times is considered. Various scaling laws of the toroidal beta with the aspect ratio are discussed. (orig.) 891 GG/orig. 892 RDG [de

Experimental study on the operating characteristics of an inner preheating transpiring wall reactor for supercritical water oxidation: Temperature profiles and product properties

International Nuclear Information System (INIS)

Zhang, Fengming; Xu, Chunyan; Zhang, Yong; Chen, Shouyan; Chen, Guifang; Ma, Chunyuan

2014-01-01

A new process to generate multiple thermal fluids by supercritical water oxidation (SCWO) was proposed to enhance oil recovery. An inner preheating transpiring wall reactor for SCWO was designed and tested to avoid plugging in the preheating section. Hot water (400–600 °C) was used as auxiliary heat source to preheat the feed to the reaction temperature. The effect of different operating parameters on the performance of the inner preheating transpiring wall reactor was investigated, and the optimized operating parameters were determined based on temperature profiles and product properties. The reaction temperature is close to 900 °C at an auxiliary heat source flow of 2.79 kg/h, and the auxiliary heat source flow is determined at 6–14 kg/h to avoid the overheating of the reactor. The useful reaction time is used to quantitatively describe the feed degradation efficiency. The outlet concentration of total organic carbon (TOC out ) and CO in the effluent gradually decreases with increasing useful reaction time. The useful reaction time needed for complete oxidation of the feed is 10.5 s for the reactor. - Highlights: • A new process to generate multiple thermal fluids by SCWO was proposed. • An inner preheating transpiring wall reactor for SCWO was designed and tested. • Hot water was used as auxiliary heat source to preheat the feed at room temperature. • Effect of operating parameters on the performance of the reactor was investigated. • The useful reaction time required for complete oxidation of the feed is 10.5 s

Atmospheric pressure flow reactor: Gas phase chemical kinetics under tropospheric conditions without wall effects

Science.gov (United States)

Koontz, Steven L. (Inventor); Davis, Dennis D. (Inventor)

1991-01-01

A flow reactor for simulating the interaction in the troposphere is set forth. A first reactant mixed with a carrier gas is delivered from a pump and flows through a duct having louvers therein. The louvers straighten out the flow, reduce turbulence and provide laminar flow discharge from the duct. A second reactant delivered from a source through a pump is input into the flowing stream, the second reactant being diffused through a plurality of small diffusion tubes to avoid disturbing the laminar flow. The commingled first and second reactants in the carrier gas are then directed along an elongated duct where the walls are spaced away from the flow of reactants to avoid wall interference, disturbance or turbulence arising from the walls. A probe connected with a measuring device can be inserted through various sampling ports in the second duct to complete measurements of the first and second reactants and the product of their reaction at selected XYZ locations relative to the flowing system.

SUITABLE LOCATION OF SHEET PILE UNDER DAM RESTING ON SANDY SOIL WITH CAVITY

Directory of Open Access Journals (Sweden)

Laith J. Aziz

2018-05-01

Full Text Available This research describes the seepage characteristics of experimental model test of dam with cutoff located at different region (at dam heel, at mid floor of dam, and at dam toe. It is resting on sandy soil with cavity at different locations in X and Y directions (such as in Al-Najaf soil city. Thirty three model tests are performed in laboratory by using steel box to estimate the quantity of the seepage and flow lines direction. It was concluded that the best location of the cutoff wall is at the dam toe for model test with cavity ( Xc B = 0 and 0.5, but for model test with cavity ( Xc B ≥1, the best location of the sheet pile wall becomes at the dam heel. For negative location of the cavity, the best location of the sheet pile wall is at the middle of the floor dam.

Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

International Nuclear Information System (INIS)

Chu, T.Y.; Slezak, S.E.; Bentz, J.H.; Pasedag, W.F.

1994-01-01

This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm 2 across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests

Experimental investigation of turbine disk cavity aerodynamics and heat transfer

Science.gov (United States)

Daniels, W. A.; Johnson, B. V.

1993-01-01

An experimental investigation of turbine disk cavity aerodynamics and heat transfer was conducted to provide an experimental data base that can guide the aerodynamic and thermal design of turbine disks and blade attachments for flow conditions and geometries simulating those of the space shuttle main engine (SSME) turbopump drive turbines. Experiments were conducted to define the nature of the aerodynamics and heat transfer of the flow within the disk cavities and blade attachments of a large scale model simulating the SSME turbopump drive turbines. These experiments include flow between the main gas path and the disk cavities, flow within the disk cavities, and leakage flows through the blade attachments and labyrinth seals. Air was used to simulate the combustion products in the gas path. Air and carbon dioxide were used to simulate the coolants injected at three locations in the disk cavities. Trace amounts of carbon dioxide were used to determine the source of the gas at selected locations on the rotors, the cavity walls, and the interstage seal. The measurements on the rotor and stationary walls in the forward and aft cavities showed that the coolant effectiveness was 90 percent or greater when the coolant flow rate was greater than the local free disk entrainment flow rate and when room temperature air was used as both coolant and gas path fluid. When a coolant-to-gas-path density ratio of 1.51 was used in the aft cavity, the coolant effectiveness on the rotor was also 90 percent or greater at the aforementioned condition. However, the coolant concentration on the stationary wall was 60 to 80 percent at the aforementioned condition indicating a more rapid mixing of the coolant and flow through the rotor shank passages. This increased mixing rate was attributed to the destabilizing effects of the adverse density gradients.

The assessment of voce coefficient for WWR-c reactor

International Nuclear Information System (INIS)

Kochnov, O.Yu.; Rybkin, N.I.

2006-01-01

The air cavity effect in WWR-ts reactor core on the total reactivity is analyzed. The experimental data of void coefficient depending on the air cavity position inside the reactor core are obtained [ru

First prototype Copper-Niobium RF Superconducting Cavity

CERN Multimedia

1983-01-01

This is the first RF superconducting cavity made of copper with a very thin layer of pure niobium deposited on the inner wall by sputtering. This new developpment lead to a considerable increase of performance and stability of superconducting cavities and to non-negligible economy. The work was carried out in the ISR workshop. This technique was adopted for the LEP II accelerating cavities. At the centre is Cristoforo Benvenuti, inventor of this important technology, with his assistants, Nadia Circelli and Max Hauer, carrying the sputtering electrode. See also 8209255, 8312339.

Injection molding of micro pillars on vertical side walls using polyether-ether-ketone (PEEK)

DEFF Research Database (Denmark)

Zhang, Yang; Hansen, Hans Nørgaard; Sørensen, Søren

2016-01-01

This paper investigates the replication of microstructures on a vertical wall by PEEK injection molding. A 4-cavity insert was used in the injection molding. Pre-fabricated nickel plates with ø 4 μm micro holes on the surface were glued on vertical walls in the cavities. 3 cavities were coated by...

Analysis of the energy transport and deposition within the reaction chamber of the Prometheus inertial fusion energy reactor

International Nuclear Information System (INIS)

Eggleston, J.E.; Abdou, M.A.; Tillack, M.S.

1995-01-01

The thermodynamic response of the Prometheus reactor chamber was analyzed and, from this analysis, a simplified thermodynamic response model was developed for parametric studies on this conceptual reactor design. This paper discusses the thermodynamic response of the cavity gas and models the condensation/evaporation of vapor to and from the first wall. Models of X-ray attenuation and ion slowing down are used to estimate the fraction of the pellet energy that is absorbed in the vapor. It was found that the gas absorbs enough energy to become partially ionized. To treat this problem, methods developed by Zel'dovich and Raizer are used in modeling the internal energy and the radiative heat flux of the vapor.From this analysis, RECON was developed, which runs with a relatively short computational time, yet retains enough accuracy for conceptual reactor design calculations. The code was used to determine whether the reactor designs could meet the stringent mass density limits that are placed on them by the physics of beam propagation through matter. RECON was also used to study the effect that the formation of a local dry spot would have on the first wall of the reactor. It was found that, for a typical reactor lifetime of 30 years, the first wall could not have a dry spot over any one section for more than 15.5 min for the laser driver design and 4.5 min for the heavy ion driver design. These times are relatively short, which implies that there is a need to keep the liquid film attached at all times. (orig.)

Development of a robot for decontamination of reactor well and maintenance pit wall surfaces

International Nuclear Information System (INIS)

Miyakawa, Minoru; Nozawa, Katsuro; Mizutani, Takeshi; Onozuka, Kazuaki; Morita, Isamu

1984-01-01

A robot has been developed at Hamaoka BWR Power Plant of Chubu Electric Power Company, Inc., which performs the decontamination of the wall surfaces of reactor wells and maintenance pits. The robot is controlled with a control box through a micro-computer. The mechanical structure and working principle of this robot is explained. One of the special features of this robot is that it perceives the steps on a wall, and washes the vertical and horizontal surfaces with two different types of brushes. As the material for the bristles of the brushes, nylon with alumina as abrasive (TAINEX-A made by Dupont Ltd.) was selected after some experience. The design specifications of the brushes were determined, based on the results of intensive performance test, which are shown in this report. The efficiency of this robot was proved by applying it to the decontamination of the reactor wells in the periodic inspection and maintenance of Unit 1 and Unit 2 in the Hamaoka BWR Power Plant. As the result of this decontamination, the contamination level was reduced from about 10 -3 μCi/cm 2 to about 10 -5 μCi/cm 2 . The measured results of contamination after the first and third decontamination works are listed for various parts of the well surfaces. (Aoki, K.)

Cavity Ring-Down Spectroscopy for Gaseous Fission Products Trace Measurements in Sodium Fast Reactors

International Nuclear Information System (INIS)

Jacquet, P.; Pailloux, A.; Doizi, D.; Aoust, G.; Jeannot, J.-P.

2013-06-01

Safety and availability are key issues of the generation IV reactors. Hence, the three radionuclide confinement barriers, including fuel cladding, must stay tight during the reactor operation. During the primary gaseous failure, fission products xenon and krypton are released. Their fast and sensitive detection guarantees the first confinement barrier tightness. In the frame of the French ASTRID project, an optical spectroscopy technique - Cavity Ring Down Spectroscopy (CRDS) - is investigated for the gaseous fission products measurement. A dedicated CRDS set-up is needed to detect the rare gases with a commercial laser. Indeed, the CRDS is coupled to a glow discharge plasma, which generates a population of metastable atoms. The xenon plasma conditions are optimized to 110 Pa and 1.3 W (3 mA). The production efficiency of metastable Xe is then 0.8 %, stable within 0.5% during hours. The metastable number density is proportional to the xenon over argon molar fraction. The spectroscopic parameters of the strong 823.16 nm xenon transition are calculated and/or measured in order to optimize the fit of the experimental spectra and make a quantitative measurement of the metastable xenon. The CRDS is coupled to the discharge cell. The laser intensity inside the cavity is limited by the optical saturation process, resulting from the strong optical pumping of the metastable state. The resulting weak CRDS signal requires a fast and very sensitive photodetector. A 600 ppt xenon molar fraction was measured by CRDS. With the present set-up, the detection limits are estimated from the baseline noise to approximately 20 ppt for each even isotope, 60 ppt for the 131 Xe and 55 ppt for the 129 Xe. This sensitivity matches the specifications required for gaseous leak measurement; approximately 100 ppt for 133 Xe (4 GBq/m 3 ) and 10 ppb for stable isotopes. The odd isotopes are selectively measured, whereas the even isotopes overlap, a spectroscopic feature that applies for stable or

Design of 118 MHz twelfth harmonic cavity of APS PAR

International Nuclear Information System (INIS)

Kang, Y.W.; Kustom, R.L.; Bridges, J.F.

1992-01-01

Two radio frequency (RF) cavities are needed in the Positron Accumulator Ring (PAR) of the Advanced Photon Source. One is for the first harmonic frequency at 9.8 MHz, and the other is for the twelfth harmonic frequency at 118 MHz. This note reports on the design of the 118 MHz RF cavity. Computer models are used to find the mode frequencies, impedances, Q-factors, and field distributions in the cavity. The computer codes MAFIA, URMEL, and URMEL-T are useful tools which model and simulate the resonance characteristics of a cavity. These codes employ the finite difference method to solve Maxwell's equations. MAFIA is a three-dimensional problem solver and uses square patches to approximate the inner surface of a cavity. URMEL and URMEL-T are two-dimensional problem solvers and use rectangular and triangular meshes, respectively. URMEL-T and MAFIA can handle problems with arbitrary dielectric materials located inside the boundary. The cavity employs a circularly cylindrical ceramic window to limit the vacuum to the beam pipe. The ceramic window used in the modeling will have a wall thickness of 0.9 cm. This wall thickness is not negligible in determining the resonant frequencies of the cavity. In the following, results of two- and three-dimensional modeling of the cavities using the URMEL-T and MAFIA codes are reported

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

Results of strategic calculations for optimizing the first wall life in a tokamak fusion reactor

International Nuclear Information System (INIS)

Daenner, W.

1981-01-01

The development of the FWLTB computer program has reached a stage where prediction of the first wall lifetime is possible. Because of the large number of free parameters strategic calculations were found to be the most appropriate way to arrive at load design conditions which allow optimum life expectancy. In this paper a revised set of life criteria is presented this being followed by the results of parameter studies in which single parameters were varied while the remaining ones were kept fixed at a reference value. These results are used as a guide during the subsequent strategic calculations. In a first strategy we aimed at finding the maximum lifetime for the case that the reactor is operated at a neutron wall loading of 10 MW/m 2 . We found that operation over a period of more than one year is possible if the first wall is designed in a very tiny geometry and cooled by a low-pressure coolant. In a second strategy the aim was to find the design conditions for the case that the first wall is cooled by a high-pressure coolant. It is shown that liquid-lithium cooling is manageable up to high wall loadings, but the lifetime is restricted to about 6 MWa/m 2 . Helium cooling allows a higher lifetime, but the design conditions are such that only modest wall loadings can be permitted. (orig.)

Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

International Nuclear Information System (INIS)

Lee, Ho Jin; Lee, B. S.; Kim, K. B.

2003-09-01

The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si 3 N 4 . Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

Condensation of ablated first-wall materials in the cascade inertial confinement fusion reactor

International Nuclear Information System (INIS)

Ladd, A.J.C.

1985-01-01

This report concerns problems involved in recondensing first-wall materials vaporized by x rays and pellet debris in the Cascade inertial confinement fusion reactor. It examines three proposed first-wall materials, beryllium oxide (BeO), silicon carbide (SiO), and pyrolytic graphite (C), paying particular attention to the chemical equilibrium and kinetics of the vaporized gases. The major results of this study are as follows. Ceramic materials composed of diatomic molecules, such as BeO and SiC, exist as highly dissociated species after vaporization. The low gas density precludes significant recombination during times of interest (i.e., less than 0.1 s). The dissociated species (Be, O, Si, and C) are, except for carbon, quite volatile and are thermodynamically stable as a vapor under the high temperature and low density found in Cascade. These materials are thus unsuitable as first-wall materials. This difficulty is avoided with pyrolytic graphite. Since the condensation coefficient of monatomic carbon vapor (approx. 0.5) is greater than that of the polyatomic vapor (<0.1), recondensation is assisted by the expected high degree of dissociation. The proposed 10-layer granular carbon bed is sufficient to condense all the carbon vapor before it penetrates to the BeO layer below. The effective condensation coefficient of the porous bed is about 50% greater than that of a smooth wall. An estimate of the mass flux leaving the chamber results in a condensation time for a carbon first wall of about 30 to 50 ms. An experiment to investigate condensation in a Cascade-like chamber is proposed

Electromagnetic design of a β=0.9, 650 MHz elliptic superconducting radio frequency cavity

International Nuclear Information System (INIS)

Jana, Arup Ratan; Kumar, V.

2011-01-01

We have recently performed two-dimensional (2D) electromagnetic design studies of a β=0.9, 650 MHz, elliptic superconducting radio frequency (SCRF) cavity using electromagnetic field solver code SUPERFISH. We have evolved the design starting from the design parameters of β=1, 1300 MHz, TESLA design SCRF cavity and then scaled it for the β=0.9 and 650 MHz case. The design has been optimized for minimizing the SCRF cavity power loss. One of the important parameters in the design of such elliptic SCRF cavities is the wall angle, which is defined as the vertical angle made by the common tangent to the iris and equator ellipses. Generally, there is a constraint on the minimum value of the wall angle, which is decided by the mechanical considerations, ease of chemical cleaning etc. In our optimization studies, we have first explored the case when there is no such constraint on wall angle. We find that from the point of view of low cavity power dissipation, the optimized design has a re-entrant geometry, where the wall angle is negative. We then perform design optimization, keeping the constraint that the wall angle should be greater than 5 degree. Keeping this constraint, we find that our optimized design parameters for the single cell match closely with the design parameters reported for Project-X. We discuss the results of 2D electromagnetic field calculations for this design using SUPERFISH. In the next, we have performed the design studies of the multi-cell β=0.9, 650 MHz, elliptic SCRF cavity. The design parameters of end-cells are optimized such that the frequency of the end-cell is matched to that of mid-cells. We have studied all the normal modes for the multi-cell cavity. The frequency of different normal modes is also calculated using a finite element code ANSYS and results are compared with those obtained using SUPERFISH. The field flatness, which is an important design criterion, is also studied. For multi-cell cavity, another important aspect is the cell

Inertial fusion reactors and magnetic fields

International Nuclear Information System (INIS)

Cornwell, J.B.; Pendergrass, J.H.

1985-01-01

The application of magnetic fields of simple configurations and modest strengths to direct target debris ions out of cavities can alleviate recognized shortcomings of several classes of inertial confinement fusion (ICF) reactors. Complex fringes of the strong magnetic fields of heavy-ion fusion (HIF) focusing magnets may intrude into reactor cavities and significantly affect the trajectories of target debris ions. The results of an assessment of potential benefits from the use of magnetic fields in ICF reactors and of potential problems with focusing-magnet fields in HIF reactors conducted to set priorities for continuing studies are reported. Computational tools are described and some preliminary results are presented

On some perculiarities of microstructure formation and the mechanical properties in thick-walled pieces of cast iron and their application as reactor structural materials

International Nuclear Information System (INIS)

Janakiev, N.

1975-01-01

The following problems are dealt with in the present work: Microstructure formation and mechanical properties of thick-walled cast pieces, influence of neutron irradiation on the mechanical properties, manufacture of thick-walled castings for reactor construction, application of cast iron as reactor structural material. It is shown that graphite formation plays an extremely important role regarding the mechanical properties. A new construction for vertically stressed pressure vessels is given. These vessels can be fabricated mainly of cast iron with graphite spheres, cast steel, or a combination of both depending on the operational pressure. (GSCH) [de

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

Containment vessel construction for nuclear power reactors

International Nuclear Information System (INIS)

Sulzer, H.D.; Coletti, J.L.

1975-01-01

A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

Development of a robot for decontamination of reactor well and maintenance pit wall surfaces

Energy Technology Data Exchange (ETDEWEB)

Miyakawa, Minoru; Nozawa, Katsuro; Mizutani, Takeshi; Onozuka, Kazuaki; Morita, Isamu

1984-03-01

A robot has been developed at Hamaoka BWR Power Plant of Chubu Electric Power Company, Inc., which performs the decontamination of the wall surfaces of reactor wells and maintenance pits. The robot is controlled with a control box through a micro-computer. The mechanical structure and working principle of this robot is explained. One of the special features of this robot is that it perceives the steps on a wall, and washes the vertical and horizontal surfaces with two different types of brushes. As the material for the bristles of the brushes, nylon with alumina as abrasive (TAINEX-A made by Dupont Ltd.) was selected after some experience. The design specifications of the brushes were determined, based on the results of intensive performance test, which are shown in this report. The efficiency of this robot was proved by applying it to the decontamination of the reactor wells in the periodic inspection and maintenance of Unit 1 and Unit 2 in the Hamaoka BWR Power Plant. As the result of this decontamination, the contamination level was reduced from 10/sup 3/ ..mu..Ci/cm/sup 2/ to about 10/sup 5/ ..mu..Ci/cm/sup 2/. The measured results of contamination after the first and third decontamination works are listed for various parts of the well surfaces.

Browns Ferry Unit 3 cavity neutron spectral analysis

International Nuclear Information System (INIS)

Martin, G.C.; Till, H.A.

1982-01-01

The General Electric Company at the Vallecitos Nuclear Center (GE-VNC) has performed neutron dosimetry measurements in the Browns Ferry Unit 3 reactor (BF3) cavity using multiple dosimeter and spectrum unfolding techniques. These measurements are the first in a BWR cavity and comprise an important part in a general program related to verification of pressure vessel integrity and to validation of calculations. Determinations of BF3 cavity neutron flux densities at five key locations at full power (1098 MWe) during core cycle 2 (November 1978 to August 1979) are presented

Plasma induced material defects and threshold values for thermal loads in high temperature resistant alloys and in refractory metals for first wall application in fusion reactors

International Nuclear Information System (INIS)

Bolt, H.; Hoven, H.; Kny, E.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1986-10-01

Materials for the application in the first wall of fusion reactors of the tokamak type are subjected to pulsed heat fluxes which range from some 0.5 MW m -2 to 10 MW m -2 during normal plasma operation, and which can exceed 1000 MW m -2 during total plasma disruptions. The structural defects and material fatigue caused by this types of plasma wall interaction are investigated and the results are plotted in threshold loading curves. Additionally, the results are, as far as possible, compared with quantitative, theoretical calculations. These procedures allow a semiquantitative evaluation of the applicability of the mentioned metals in the first wall of fusion reactors. (orig.) [de

PEP-II RF cavity revisited

International Nuclear Information System (INIS)

Rimmer, R.A.; Koehler, G.; Li, D.; Hartman, N.; Folwell, N.; Hodgson, J.; Ko, K.; McCandless, B.

1999-01-01

This report describes the results of numerical simulations of the PEP-II RF cavity performed after the completion of the construction phase of the project and comparisons are made to previous calculations and measured results. These analyses were performed to evaluate new calculation techniques for the HOM distribution and RF surface heating that were not available at the time of the original design. These include the use of a high frequency electromagnetic element in ANSYS and the new Omega 3P code to study wall losses, and the development of broadband time domain simulation methods in MAFIA for the HOM loading. The computed HOM spectrum is compared with cavity measurements and observed beam-induced signals. The cavity fabrication method is reviewed, with the benefit of hindsight, and simplifications are discussed

Effect of Sweep on Cavity Flow Fields at Subsonic and Transonic Speeds

Science.gov (United States)

Tracy, Maureen B.; Plentovich, Elizabeth B.; Hemsch, Michael J.; Wilcox, Floyd J.

2012-01-01

An experimental investigation was conducted in the NASA Langley 7 x 10-Foot High Speed Tunnel (HST) to study the effect of leading- and trailing-edge sweep on cavity flow fields for a range of cavity length-to-height (l/h) ratios. The free-stream Mach number was varied from 0.2 to 0.8. The cavity had a depth of 0.5 inches, a width of 2.5 inches, and a maximum length of 12.0 inches. The leading- and trailing-edge sweep was adjusted using block inserts to achieve leading edge sweep angles of 65 deg, 55 deg, 45 deg, 35 deg, and 0 deg. The fore and aft cavity walls were always parallel. The aft wall of the cavity was remotely positioned to achieve a range of length-to-depth ratios. Fluctuating- and static-pressure data were obtained on the floor of the cavity. The fluctuating pressure data were used to determine whether or not resonance occurred in the cavity rather than to provide a characterization of the fluctuating pressure field. Qualitative surface flow visualization was obtained using a technique in which colored water was introduced into the model through static-pressure orifices. A complete tabulation of the mean static-pressure data for the swept leading edge cavities is included.

Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jin; Lee, B. S.; Kim, K. B

2003-09-01

The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

Chamber and Wall Response to Target Implosion in Inertial and Z-Pinch Fusion and Lithography Devices

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.; Morozov, V.; Sizyuk, V.

2006-01-01

The chamber walls, both solid and liquid, in inertial fusion energy (IFE) and Z-pinch reactors and Lithography devices are exposed to harsh conditions following each target implosion or pinching of plasma. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to target implosion. Detailed models have been developed for reflected laser light, emitted photons, neutrons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The hydrodynamic response of chamber walls in bare or in gas-filled cavities and the photon transport of the deposited energy has been calculated by means of new and advanced numerical techniques for accurate shock treatment and propagation. These models include detail media hydrodynamics, non-LTE multi-group for both continuum and line radiation transport, and dynamics of eroded debris resulting from the intense energy deposition. The focus of this study is to critically assess the reliability and the dynamic response of chamber walls in various proposed protection methods for IFE systems. Key requirements are that: (i) the chamber wall accommodates the cyclic energy deposition while providing the required lifetime due to various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls, and (ii) after each shot the chamber is cleared and returned to a quiescent state in preparation for the target injection and the firing of the driver for the subsequent shot. This paper investigates in details these two important issues and found that the required operating frequency of the IFE reactors for power production may be severely limited due to these two requirements. (author)

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Impact of Lorentz forces on a Spoke cavity with β 0.15 and on a Spiral-2 cavity with β 0.12

International Nuclear Information System (INIS)

Gassot, H.

2007-01-01

Mono-spoke superconducting cavities have been proposed for the acceleration of radioactive ion beams. The interaction of the electromagnetic field with the surface electrical current generates Lorentz forces that operate on the intern wall of the cavity, the distribution of these forces is highly non-linear and varying. The stability of a superconducting cavity is directly linked to the frequency variation due to Lorentz forces and as a consequence the optimized design of a cavity must take into account these forces. In order to optimize the design of a cavity, 3 complementary software have been developed: Catia, a computer-aided-design software, Soprano for electromagnetic modeling and Cast3m for mechanical modeling. Preliminary results show a good agreement between predicted values and experimental data. (A.C.)

Thermodynamic modelling and solar reactor design for syngas production through SCWG of algae

Science.gov (United States)

Venkataraman, Mahesh B.; Rahbari, Alireza; Pye, John

2017-06-01

Conversion of algal biomass into value added products, such as liquid fuels, using solar-assisted supercritical water gasification (SCWG) offers a promising approach for clean fuel production. SCWG has significant advantages over conventional gasification in terms of flexibility of feedstock, faster intrinsic kinetics and lower char formation. A relatively unexplored avenue in SCWG is the use of non-renewable source of energy for driving the endothermic gasification. The use of concentrated solar thermal to provide the process heat is attractive, especially in the case of expensive feedstocks such as algae. This study attempts to identify the key parameters and constraints in designing a solar cavity receiver/reactor for on-sun SCWG of algal biomass. A tubular plug-flow reactor, operating at 24 MPa and 400-600 °C with a solar input of 20MWth is modelled. Solar energy is utilized to increase the temperature of the reaction medium (10 wt.% algae solution) from 400 to 605 °C and simultaneously drive the gasification. The model additionally incorporates material constraints based on the allowable stresses for a commercially available Ni-based alloy (Inconel 625), and exergy accounting for the cavity reactor. A parametric evaluation of the steady state performance and quantification of the losses through wall conduction, external radiation and convection, internal convection, frictional pressure drop, mixing and chemical irreversibility, is presented.

Report of the study meeting on the interaction between plasma and the first wall of a fusion reactor

International Nuclear Information System (INIS)

Miyahara, Akira; Akaishi, Kenya; Kawamura, Takaichi; Kabetani, Zenzaburo; Sagara, Akio.

1978-12-01

The study meeting on the interaction between plasma and the first wall of a fusion reactor was held from July 24 to July 27, 1978. At this meeting, discussions were made on the interaction between plasma and wall and the effect of impurities. Reports on the ISS observation concerning the Mo surface as a limiter, on the measurement of sputter rate by a microbalance, on the surface roughness of the materials for the first wall at the atomic order, on the selective sputtering of binary alloys, and on the physical and chemical sputtering on the material surface of C and SiC were also presented. The research projects of the Institute of Plasma Physics and Hokkaido University were introduced. Collaboration of two groups was considered. (Kato, T.)

Water-immersion type ship reactor

International Nuclear Information System (INIS)

Okada, Hiroki; Yamamura, Toshio.

1996-01-01

In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

Mixed convection in a lid-driven square cavity with partial slip

International Nuclear Information System (INIS)

Ismael, Muneer A.; Pop, Ioan; Chamkha, Ali J.

2014-01-01

Steady laminar mixed convection inside a lid-driven square cavity filled with water is studied numerically. The lid is due to the movement of the isothermal top and bottom walls which are maintained at T c and T h , respectively, with T h is higher than T c . A partial slip condition was imposed in these two moving walls. The vertical walls of the cavity are kept adiabatic. The appliance of the numerical analysis was USR finite difference method with upwind scheme treatments of the convective terms included in the momentum and energy equations. The studied relevant parameters were: the partial slip parameter S (0-∞); Richardson number Ri (0.01-100) and the direction of the moving walls (λ t = 1, λ b = ±1). The results have showed that there are critical values for the partial slip parameter at which the convection is declined. (authors)

Numerical study of free convection in an enclosure with two vertical isothermal walls

International Nuclear Information System (INIS)

Barletta, A.; Rossi di Schio, E.; Zanchini, E.; Nobile, E.; Pinto, F.

2005-01-01

In this paper, natural convection is studied in a 2D-cavity with two vertical isothermal walls, kept at different temperatures, and two adiabatic walls which are either straight (rectangular cavity) or elliptic (modified rectangular cavity). The local mass, momentum and energy balance equations are written in a dimensionless form and solved numerically, by means of two different software packages based on Galerkin finite element methods. With reference to a Prandtl number of 0.71, two rectangular cavities are studied: a square one and a cavity with height double than width. Then, for each value of the ratio between height and width, two cavities with elliptic boundaries are investigated. The numerical solution shows that the elliptic boundaries enhance the mean Nusselt number and the dimensionless mean kinetic energy of the fluid. (authors)

Upgrade of the Annular Core Pulse Reactor

Energy Technology Data Exchange (ETDEWEB)

Reuscher, J A [Sandia Laboratories, Albuquerque, NM (United States)

1976-07-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past two years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 by utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. Preliminary studies have identified several potential approaches to the ACPR performance improvement. The most promising approach appears to be the two-region core concept. The inner region, surrounding the irradiation cavity, would consist of a high-heat capacity fuel capable of absorbing the fission energy associated with a large nuclear pulse. The number of fissions occurring near the cavity would be greatly increased which, in turn, would significantly increase the fluence in the cavity. The outer region would consist of a U-ZrH fuel to provide an overall negative temperature coefficient for the two region core. Two candidate high heat capacity fuels [(BeO-UO{sub 2} and UC-ZrC) - graphite] are under consideration. Since this reactor upgrade represents a modification to an existing facility, it can be achieved in a relatively short time. It is anticipated that most of the existing reactor structure can be used for the upgrade. The present core occupies about one-half of the location in the grid plate. The high-heat capacity fuel

Mixed convection of ferrofluids in a lid driven cavity with two rotating cylinders

Directory of Open Access Journals (Sweden)

Fatih Selimefendigil

2015-09-01

Full Text Available Mixed convection of ferrofluid filled lid driven cavity in the presence of two rotating cylinders were numerically investigated by using the finite element method. The cavity is heated from below, cooled from driven wall and rotating cylinder surfaces and side vertical walls of the cavity are assumed to be adiabatic. A magnetic dipole source is placed below the bottom wall of the cavity. The study is performed for various values of Reynolds numbers (100 ≤ Re ≤ 1000, angular rotational speed of the cylinders (−400 ≤ Ω ≤ 400, magnetic dipole strengths (0 ≤ γ ≤ 500, angular velocity ratios of the cylinders (0.25≤Ωi/Ωj≤4 and diameter ratios of the cylinders (0.5≤Di/Dj≤2. It is observed that flow patterns and thermal transport within the cavity are affected by variation in Reynolds number and magnetic dipole strength. The results of this investigation revealed that cylinder angular velocities, ratio of the angular velocities and diameter ratios have profound effect on heat transfer enhancement within the cavity. Averaged heat transfer enhancements of 181.5 % is achieved for clockwise rotation of the cylinder at Ω = −400 compared to motionless cylinder case. Increasing the angular velocity ratio from Ω2/Ω1=0.25 to Ω2/Ω1=4 brings about 91.7 % of heat transfer enhancement.

Cooling device for thermonuclear reactor and modular packing block for the wall realization of a such device

International Nuclear Information System (INIS)

Archer, J.; Stalport, G.; Besson, D.; Faron, R.; Coulon, M.

1988-01-01

The cooling device for a thermonuclear reactor wall is made by modular thermally conductive heat-resistant blocks (graphite by example), a prismatic head on one face of each block, the opposite face bearing against cooling tubes, a base to each block with an aperture and rods passing through the apertures reversibly fixing each row of blocks to a support [fr

Reactor container

International Nuclear Information System (INIS)

Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

1979-01-01

Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

Effects of Core Cavity on a Flow Distribution

Energy Technology Data Exchange (ETDEWEB)

Kwon, Tae-Soon; Kim, Kihwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The axial pressure drop is removed in the free core condition, But the actual core has lots of fuel bundles and mixing vanes to the flow direction. The axial pressure drop induces flow uniformity. In a uniform flow having no shear stress, the cross flow or cross flow mixing decreases. The mixing factor is important in the reactor safety during a Steam Line Break (SLB) or Main Steam Line Break (MSLB) transients. And the effect of core cavity is needed to evaluate the realistic core mixing factor quantification. The multi-dimensional flow mixing phenomena in a core cavity has been studied using a CFD code. The 1/5-scale model was applied for the reactor flow analysis. A single phase water flow conditions were considered for the 4-cold leg and DVI flows. To quantify the mixing intensity, a boron scalar was introduced to the ECC injection water at cold legs and DVI nozzles. The present CFD pre-study was performed to quantify the effects of core structure on the mixing phenomena. The quantified boron mixing scalar in the core simulator model represented the effect of core cavity on the core mixing phenomena. This simulation results also give the information for sensor resolution to measure the boron concentration in the experiments and response time to detect mixing phenomena at the core and reactor vessel.

A model for the computation of the thermal processes in the reactor cavity during a severe accident in a LWR, at the presence of sump water, from the time of reactor pressure vessel failure to the start time of melt/concrete interaction

International Nuclear Information System (INIS)

Hirschmann, H.

1990-04-01

At present no experimental results are available which analyze that stage of a severe accident in a light water reactor, during which the reactor pressure vessel fails by melting, the core debris relocates into the water pool on the floor of the containment building (cavity) and again is heated up. Therefore an analytical model is described, with the help of which the process of material relocation, the heating of the material in the cavity interacting with the pool water, and the production rates of vapour and hydrogen can be estimated. The slumped mass accumulating in the cavity is taken to be the sum of infinitely small mass parts, assumed to slump at different times, which after slumping undergo individual thermal histories. The enthalpy of the slumped mass is the sum of the enthalpies of the single mass parts. The average temperature of the slumped mass is given by the enthalpy computed in this manner. The production rates of the gases are additive superpositions of all partial rates from the mass parts. The gas rates are computed using the balance of enthalpy and mass. (author) 5 refs

Development of fatigue life criteria for experimental fusion reactor first-wall structures

International Nuclear Information System (INIS)

Nickell, R.E.; Esztegar, E.P.

1980-01-01

An approach to the rational design of fusion reactor first-wall structures against fatigue crack growth is proposed. The approach is motivated by microstructural observations of fatigue crack growth enhancement in uniruniradiated materials due to volumetric damage ahead of a propagating crack. Examples are cited that illustrate the effect of mean stress on void nucleation and coalescence, which represent the dominant form of volumetric damage at low temperature, and of grain boundary sliding and creep cavitation, which are the dominant volumetric damage mechanisms at high temperature. The analogy is then drawn between these forms of fatigue crack growth enhancement and those promoted by irradiation exposure in the fusion reactor environment, such as helium embrittlement and atomic displacement. An enhanced strain range is suggested as a macroscopic measure of the reduction in fatigue life due to the higher fatigue crack growth rates. The enhanced strain range permits a separation of volumetric and cyclic effects, and assists in the assignment of rational design factors to each effect. A series of experiments are outlined which should provide the numerical values of the parameters for the enhanced strain range. (orig.)

Numerical Investigation of Merged and Non-merged Flame of a Twin Cavity Annular Trapped Vortex Combustor

Directory of Open Access Journals (Sweden)

Pravendra Kumar

2016-09-01

Full Text Available : The present work is focused to characterize numerically the merged and non-merged flame emanating from the cavities in downstream of twin cavity Annular Trapped Vortex Combustor (ATVC.The isotherm corresponding to the auto-ignition temperature is used to locate the merging point of the flame in the mainstream region along the combustor length. In present study, the cavity flame is said to be merged only if this isotherm corresponding to self-ignition temperature of methane is located within 20 percentage of the combustor length from aft wall of cavities. It is interesting to note that on increasing the power loading parameter (PLP in mainstream for a constant power loading parameter ratio (outer to inner cavity, the merging point gets shifted towards the cavity aft-wall. This leads to the reduction of combustor length and subsequent reduction in overall weight of the gas turbine engine.

The use of microperforated plates to attenuate cavity resonances

DEFF Research Database (Denmark)

Fenech, Benjamin; Keith, Graeme; Jacobsen, Finn

2006-01-01

The use of microperforated plates to introduce damping in a closed cavity is examined. By placing a microperforated plate well inside the cavity instead of near a wall as traditionally done in room acoustics, high attenuation can be obtained for specific acoustic modes, compared with the lower...... attenuation that can be obtained in a broad frequency range with the conventional position of the plate. An analytical method for predicting the attenuation is presented. The method involves finding complex eigenvalues and eigenfunctions for the modified cavity and makes it possible to predict Green......'s functions. The results, which are validated experimentally, show that a microperforated plate can provide substantial attenuation of modes in a cavity. One possible application of these findings is the treatment of boiler tones in heat-exchanger cavities....

Effect of Perpendicular Magnetic Field on Free Convection in a Rectangular Cavity

Directory of Open Access Journals (Sweden)

Anand Kumar

2015-12-01

Full Text Available The steady free convective flow of a viscous incompressible and electrically conducting fluid in a two-dimensional cavity in the presence of a magnetic field applied normal to the plane of the cavity is investigated. The side vertical walls of the cavity are heated differentially while the horizontal walls are assumed to be insulated. The governing equations are re-formulated in terms of vorticity and stream function. The resulting boundary value problem is solved numerically using an alternating direction implicit (ADI method. A number of plots illustrating the influence of Hartmann number and Rayleigh number on the streamlines and isotherms as well as the velocity and temperature profiles are shown. Furthermore, results for the average Nusselt number and the maximum absolute stream function have been obtained, and these are compared with the corresponding results in the literature when the magnetic field is applied along the cavity in the horizontal direction.

Three-Dimensional Numerical Evaluation of Thermal Performance of Uninsulated Wall Assemblies: Preprint

Energy Technology Data Exchange (ETDEWEB)

Ridouane, E. H.; Bianchi, M.

2011-11-01

This study describes a detailed three-dimensional computational fluid dynamics modeling to evaluate the thermal performance of uninsulated wall assemblies accounting for conduction through framing, convection, and radiation. The model allows for material properties variations with temperature. Parameters that were varied in the study include ambient outdoor temperature and cavity surface emissivity. Understanding the thermal performance of uninsulated wall cavities is essential for accurate prediction of energy use in residential buildings. The results can serve as input for building energy simulation tools for modeling the temperature dependent energy performance of homes with uninsulated walls.

Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

International Nuclear Information System (INIS)

Swami, H.L.; Danani, C.

2015-01-01

In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

Post caesarean section anterior abdominal wall endometriosis ...

African Journals Online (AJOL)

Abdominal wall endometriosis is a likely sequelae of caesarean section as viable endometrial tissue are deposited in the peritoneal cavity or anterior abdominal wall. One such case to sensitize clinicians of this rare presentation of the disease is presented. The patient was a 48 year old woman who presented with a lesion ...

Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

Science.gov (United States)

Woolley, Robert D.

2002-01-01

A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

Integrated Life Cycle Energy and Greenhouse Gas Analysis of Exterior Wall Systems for Residential Buildings

Directory of Open Access Journals (Sweden)

Reza Broun

2014-11-01

Full Text Available This paper investigates the breakdown of primary energy use and greenhouse gas (GHG emissions of two common types of exterior walls in the U.K.: insulated concrete form (ICF and cavity walls. A comprehensive assessment was conducted to evaluate the environmental performance of each exterior wall system over 50 years of service life in Edinburgh and Bristol. The results indicate that for both wall systems, use phase is the major contributor to the overall environmental impacts, mainly due to associated electricity consumption. For the ICF wall system in Edinburgh, 91% of GHG emissions were attributed to the use phase, with 7.8% in the pre-use and 1.2% in end-of-life phases. For the same system in Bristol, emissions were 89%, 9% and 2%, respectively. A similar trend was observed for cavity wall systems in both locations. It was concluded that in each scenario, the ICF wall system performed better when compared to the cavity wall system. The results of the sensitivity analysis clearly show that the uncertainties relevant to the change of the thickness of the wall are quite tolerable: variable up to 5%, as far as energy and greenhouse emissions are concerned.

Buoyancy Induced Heat Transfer and Fluid Flow Inside a Prismatic Cavity

International Nuclear Information System (INIS)

Aich, Walid; Omri, Ahmed; Ben Nasrallah, Sassi

2009-01-01

This paper deals with a numerical simulation of natural convection flows in a prismatic cavity. This configuration represents solar energy collectors, conventional attic spaces of greenhouses and buildings with pitched roofs. The third dimension of the cavity is considered long enough for the flow to be considered 2D. The base is submitted to a uniform heat flux, the two top inclined walls are symmetrically cooled and the two vertical walls are assumed to be perfect thermal insulators. The aim of the study is to examine the thermal exchange by natural convection and effects of buoyancy forces on flow structure. The study provides useful information on the flow structure sensitivity to the governing parameters, the Rayleigh number (Ra) and the aspect ratio of the cavity. The hydrodynamic and thermal fields, the local Nusselt number, the temperature profile at the bottom and at the center of the cavity are investigated for a large range of Ra. The effect of the aspect ratio is examined for different values of Ra. Based on the authors knowledge, no previous results on natural convection in this geometry exist

Mixed convection in inclined lid driven cavity by Lattice Boltzmann Method and heat flux boundary condition

International Nuclear Information System (INIS)

D'Orazio, A; Karimipour, A; Nezhad, A H; Shirani, E

2014-01-01

Laminar mixed convective heat transfer in two-dimensional rectangular inclined driven cavity is studied numerically by means of a double population thermal Lattice Boltzmann method. Through the top moving lid the heat flux enters the cavity whereas it leaves the system through the bottom wall; side walls are adiabatic. The counter-slip internal energy density boundary condition, able to simulate an imposed non zero heat flux at the wall, is applied, in order to demonstrate that it can be effectively used to simulate heat transfer phenomena also in case of moving walls. Results are analyzed over a range of the Richardson numbers and tilting angles of the enclosure, encompassing the dominating forced convection, mixed convection, and dominating natural convection flow regimes. As expected, heat transfer rate increases as increases the inclination angle, but this effect is significant for higher Richardson numbers, when buoyancy forces dominate the problem; for horizontal cavity, average Nusselt number decreases with the increase of Richardson number because of the stratified field configuration

Microstructural evolution in an austenitic stainless steel fusion reactor first wall

International Nuclear Information System (INIS)

Stoller, R.E.; Odette, G.R.

1986-01-01

A detailed rate-theory-based model of microstructural evolution under fast neutron irradiation has been developed. The prominent new aspect of this model is a treatment of dislocation evolution in which Frank faulted loops nucleate, grow and unfault to provide a source for network dislocations while the dislocation network can be simultaneously annihilated by a climb/glide process. The predictions of this model compare very favorably with the observed dose and temperature dependence of these key microstructural features over a broad range. This new description of dislocation evolution has been coupled with a previously developed model of cavity evolution and good agreement has been obtained between the predictions of the composite model and fast reactor swelling data as well. The results from the composite model also reveal that the various components of the irradiation-induced microstructure evolve in a highly coupled manner. The predictions of the composite model are more sensitive to parametric variations than more simple models. Hence, its value as a tool in data analysis and extrapolation is enhanced

Application of backscatter electrons for large area imaging of cavities produced by neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Pastukhov, V.I. [Joint Stock Company “Institute of Nuclear Materials” (JSC “INM”), Zarechny, Sverdlovsk Region (Russian Federation); Ural Federal University Named After the First President of Russia, B. N. Yeltsyn, Ekaterinburg (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow (Russian Federation); Averin, S.A.; Panchenko, V.L. [Joint Stock Company “Institute of Nuclear Materials” (JSC “INM”), Zarechny, Sverdlovsk Region (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow (Russian Federation); Portnykh, I.A. [Joint Stock Company “Institute of Nuclear Materials” (JSC “INM”), Zarechny, Sverdlovsk Region (Russian Federation); Freyer, P.D. [Westinghouse Electric Company, Pittsburgh, PA (United States); Giannuzzi, L.A. [L.A. Giannuzzi & Associates LLC, Fort Myers, FL (United States); Garner, F.A., E-mail: frank.garner@dslextreme.com [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow (Russian Federation); Radiation Effects Consulting LLC, Richland, WA (United States); Texas A& M University, College Station, TX (United States)

2016-11-15

It is shown that with proper optimization, backscattered electrons in a scanning electron microscope can produce images of cavity distribution in austenitic steels over a large specimen surface for a depth of ∼500–700 nm, eliminating the need for electropolishing or multiple specimen production. This technique is especially useful for quantifying cavity structures when the specimen is known or suspected to contain very heterogeneous distributions of cavities. Examples are shown for cold-worked EK-164, a very heterogeneously-swelling Russian fast reactor fuel cladding steel and also for AISI 304, a homogeneously-swelling Western steel used for major structural components of light water cooled reactors. This non-destructive overview method of quantifying cavity distribution can be used to direct the location and number of required focused ion beam prepared transmission electron microscopy specimens for examination of either neutron or ion-irradiated specimens. This technique can also be applied in stereo mode to quantify the depth dependence of cavity distributions.

Heat and mass transfer in porous cavity: Assisting flow

Energy Technology Data Exchange (ETDEWEB)

Badruddin, Irfan Anjum [Dept. of Mechanical Engineering, University of Malaya, Kuala Lumpur, 50603 (Malaysia); Quadir, G. A. [School of Mechatronic Engineering, University Malaysia Perlis, Pauh Putra, 02600 Arau, Perlis (Malaysia)

2016-06-08

In this paper, investigation of heat and mass transfer in a porous cavity is carried out. The governing partial differential equations are non-dimensionalised and solved using finite element method. The left vertical surface of the cavity is maintained at constant temperature and concentration which are higher than the ambient temperature and concentration applied at right vertical surface. The top and bottom walls of the cavity are adiabatic. Heat transfer is assumed to take place by natural convection and radiation. The investigation is carried out for assisting flow when buoyancy and gravity force act in same direction.

Experimental study of bypass flow in near wall gaps of a pebble bed reactor using hot wire anemometry technique

International Nuclear Information System (INIS)

Amini, Noushin; Hassan, Yassin A.

2014-01-01

Highlights: • Coolant flow behavior in near wall gaps of a pebble bed reactor is studied. • Hot wire anemometry is applied for high frequency velocity measurements. • Bypass flow is identified within the velocity profiles of near wall gaps. • Effect of gap geometry and Reynolds number on bypass flow is investigated. • Variation of velocity power spectra with radial location and Reynolds number is studied. - Abstract: Coolant flow behavior through the core of an annular pebble bed reactor is investigated in this experimental study. A high frequency hot wire anemometry system coupled with an X-probe is used for measurement of axial and radial velocity components at different points within two near wall gaps at five different modified Reynolds numbers (Re m = 2043–6857). The velocity profiles within the gaps verify the presence of an area of increased velocity close to the pebble bed outer reflector wall, which is known as the bypass flow. Moreover, the characteristics of the coolant flow profile are seen to be highly dependent on the gap geometry. The effect of Reynolds number on the velocity profiles varies as the geometry of the gap changes. The time histories of the local velocities measured with considerably high frequency are further analyzed using power spectral density technique. Power spectral plots illustrate substantial spatial variation of the energy content, spectral shape, and the slope of the energy cascade region. A significant correlation between Reynolds number and characteristics of the velocity power spectra is observed

Auxiliary reactor for a hydrocarbon reforming system

Science.gov (United States)

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

2006-01-17

An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

Improving the efficiency of microwave devices with a double output cavity

International Nuclear Information System (INIS)

Eppley, K.R.; Herrmannsfeldt, W.B.; Lee, T.G.

1986-05-01

Double output cavities have been used experimentally to increase the efficiency of high-power klystrons. We have used particle-in-cell simulations with the 2 + 1/2 dimensional code MASK to optimize the design of double output cavities for the lasertron and the 50 MW klystron under development at SLAC. We discuss design considerations for double output cavities (e.g., optimum choice of voltages and phases, efficiency, wall interception, breakdown). We describe how one calculates the cavity impedance matrix from the gap voltages and phases. Simulation results are compared to experience with the 150 MW klystron

Implosion of Cylindrical Cavities via Short Duration Impulsive Loading

Science.gov (United States)

Huneault, Justin; Higgins, Andrew

2014-11-01

An apparatus has been developed to study the collapse of a cylindrical cavity in gelatin subjected to a symmetric impact-driven impulsive loading. A gas-driven annular projectile is accelerated to approximately 50 m/s, at which point it impacts a gelatin casting confined by curved steel surfaces that allow a transition from an annular geometry to a cylindrically imploding motion. The implosion is visualized by a high-speed camera through a window which forms the top confining wall of the implosion cavity. The initial size of the cavity is such that the gelatin wall is two to five times thicker than the impacting projectile. Thus, during impact the compression wave which travels towards the cavity is closely followed by a rarefaction resulting from the free surface reflection of the compression wave in the projectile. As the compression wave in the gelatin reaches the inner surface, it will also reflect as a rarefaction wave. The interaction between the rarefaction waves from the gelatin and projectile free surfaces leads to large tensile stresses resulting in the spallation of a relatively thin shell. The study focuses on the effect of impact parameters on the thickness and uniformity of the imploding shell formed by the cavitation in the imploding gelatin cylinder.

The thermal response of the first wall of a fusion reactor blanket to plasma disruptions

International Nuclear Information System (INIS)

Klippel, H.Th.

1983-09-01

Major plasma disruptions in Tokamak power reactors are potentially dangerous because high thermal overloading of the first wall may occur, resulting in melting and evaporation. The present uncertainties of the disruption characteristics, in particular the space and time dependence of the energy deposition, lead to a wide variation in the prospective surface energy loads. The thermal response of a first wall of aluminium, stainless steel and of graphite subjected to disruption energy loads up to 1000 J cm -2 has been analysed including the effects of melting and surface evaporation, vapour recondensation, vapour shielding, and the moving of the surface boundary caused by the evaporation. A special calculation model has been developed for this purpose. The main results are the following: by values of local transient energy depositions over 1500 J cm -2 bare stainless steel walls are damaged severely. Further calculations are needed to estimate the endurance limit of several candidate first wall materials. Applications of coatings on surfaces need special attention. For the reference INTOR disruption (approx. 100 J cm -2 ) evaporation is not significant. The effect of vapour shielding on evaporation has been found to be significant. The effect on melting is less pronounced. In a complete analysis the stability and dynamic behaviour of the melted layer under electromagnetic forces should be included. Also a reliable set of plasma disruption characteristics should be gathered

Lattice Boltzmann simulations of three-dimensional incompressible flows in a four-sided lid driven cavity

Energy Technology Data Exchange (ETDEWEB)

Li, Cheng Gong [National Engineering Laboratory for MTO, Dalian National Laboratory for Clean Energy, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); Maa, Jerome P-Y, E-mail: chenggongli@dicp.ac.cn [Virginia Institute of Marine Science, College of William and Mary, Gloucester Point, VA 23062 (United States)

2017-04-15

Numerical study on three-dimensional (3D), incompressible, four-sided lid (FSL) driven cavity flows has been conducted to show the effects of the transverse aspect ratio, K , on the flow field by using a multiple relaxation time lattice Boltzmann equation. The top wall is driven from left to right, the left wall is moved downward, whereas the right wall is driven upward, and the bottom wall is moved from right to left, all the four moving walls have the same speed and the others boundaries are fixed. Numerical computations are performed for several Reynolds numbers for laminar flows, up to 1000, with various transverse aspect ratios. The flow can reach a steady state and the flow pattern is symmetric with respect to the two cavity diagonals (i.e., the center of the cavity). At Reynolds number = 300, the flow structures of the 3D FSL cavity flow at steady state with various transverse aspect ratio, i.e., 3, 2, 1, 0.75, 0.5 and 0.25 only show the unstable symmetrical flow pattern. The stable asymmetrical flow pattern could be reproduced only by increasing the Reynolds number that is above a critical value which is dependent on the aspect ratio. It is found that an aspect ratio of more than 5 is needed to reproduce flow patterns, both symmetric and asymmetric flows, simulated by using 2D numerical models. (paper)

Stable–streamlined and helical cavities following the impact of Leidenfrost spheres

KAUST Repository

Mansoor, Mohammad M.

2017-06-23

We report results from an experimental study on the formation of stable–streamlined and helical cavity wakes following the free-surface impact of Leidenfrost spheres. Similar to the observations of Mansoor et al. (J. Fluid Mech., vol. 743, 2014, pp. 295–326), we show that acoustic ripples form along the interface of elongated cavities entrained in the presence of wall effects as soon as the primary cavity pinch-off takes place. The crests of these ripples can act as favourable points for closure, producing multiple acoustic pinch-offs, which are found to occur in an acoustic pinch-off cascade. We show that these ripples pacify with time in the absence of physical contact between the sphere and the liquid, leading to extremely smooth cavity wake profiles. More importantly, the downward-facing jet at the apex of the cavity is continually suppressed due to a skin-friction drag effect at the colliding cavity-wall junction, which ultimately produces a stable–streamlined cavity wake. This streamlined configuration is found to experience drag coefficients an order of a magnitude lower than those acting on room-temperature spheres. A striking observation is the formation of helical cavities which occur for impact Reynolds numbers and are characterized by multiple interfacial ridges, stemming from and rotating synchronously about an evident contact line around the sphere equator. The contact line is shown to result from the degeneration of Kelvin–Helmholtz billows into turbulence which are observed forming along the liquid–vapour interface around the bottom hemisphere of the sphere. Using sphere trajectory measurements, we show that this helical cavity wake configuration has 40 %–55 % smaller force coefficients than those obtained in the formation of stable cavity wakes.

Stable–streamlined and helical cavities following the impact of Leidenfrost spheres

KAUST Repository

Mansoor, Mohammad M.; Vakarelski, Ivan Uriev; Marston, J. O.; Truscott, T. T.; Thoroddsen, Sigurdur T

2017-01-01

We report results from an experimental study on the formation of stable–streamlined and helical cavity wakes following the free-surface impact of Leidenfrost spheres. Similar to the observations of Mansoor et al. (J. Fluid Mech., vol. 743, 2014, pp. 295–326), we show that acoustic ripples form along the interface of elongated cavities entrained in the presence of wall effects as soon as the primary cavity pinch-off takes place. The crests of these ripples can act as favourable points for closure, producing multiple acoustic pinch-offs, which are found to occur in an acoustic pinch-off cascade. We show that these ripples pacify with time in the absence of physical contact between the sphere and the liquid, leading to extremely smooth cavity wake profiles. More importantly, the downward-facing jet at the apex of the cavity is continually suppressed due to a skin-friction drag effect at the colliding cavity-wall junction, which ultimately produces a stable–streamlined cavity wake. This streamlined configuration is found to experience drag coefficients an order of a magnitude lower than those acting on room-temperature spheres. A striking observation is the formation of helical cavities which occur for impact Reynolds numbers and are characterized by multiple interfacial ridges, stemming from and rotating synchronously about an evident contact line around the sphere equator. The contact line is shown to result from the degeneration of Kelvin–Helmholtz billows into turbulence which are observed forming along the liquid–vapour interface around the bottom hemisphere of the sphere. Using sphere trajectory measurements, we show that this helical cavity wake configuration has 40 %–55 % smaller force coefficients than those obtained in the formation of stable cavity wakes.

Slot-coupled CW standing wave accelerating cavity

Energy Technology Data Exchange (ETDEWEB)

Wang, Shaoheng; Rimmer, Robert; Wang, Haipeng

2017-05-16

A slot-coupled CW standing wave multi-cell accelerating cavity. To achieve high efficiency graded beta acceleration, each cell in the multi-cell cavity may include different cell lengths. Alternatively, to achieve high efficiency with acceleration for particles with beta equal to 1, each cell in the multi-cell cavity may include the same cell design. Coupling between the cells is achieved with a plurality of axially aligned kidney-shaped slots on the wall between cells. The slot-coupling method makes the design very compact. The shape of the cell, including the slots and the cone, are optimized to maximize the power efficiency and minimize the peak power density on the surface. The slots are non-resonant, thereby enabling shorter slots and less power loss.

Lateral restraint assembly in a nuclear reactor

International Nuclear Information System (INIS)

Brown, S.J.; Gorholt, W.

1977-01-01

A lateral restraint assembly is described for a reactor of, for example, the high temperature gas-cooled type which commonly includes a reactor core of relatively complex construction supported within a shell or vessel providing a shielded cavity for containing the reactor core. (U.K.)

Developments in modelling the effect of aerosol on the thermal performance of the Fast Reactor cover gas space

International Nuclear Information System (INIS)

Ford, I.J.; Clement, C.F.

1990-03-01

The sodium aerosol which forms in the cover gas space of a Fast Reactor couples the processes of heat and mass transfer to and from the bounding surfaces and affects the thermal performance of the cavity. This report describes extensions to previously separate models of heat transfer and aerosol formation and removal in the cover gas space, and the linking of the two calculations in a consistent manner. The extensions made to the theories include thermophoretic aerosol removal, radiative-driven redistribution in aerosol sizes, and the side-wall influence on the bulk cavity temperature. The link between aerosol properties and boundary layer saturations is also examined, especially in the far-from-saturated limit. The models can be used in the interpretation of cover gas space experiments and some example calculations are given. (author)

Design of half-reentrant SRF cavities

International Nuclear Information System (INIS)

Meidlinger, M.; Grimm, T.L.; Hartung, W.

2006-01-01

The shape of a TeSLA inner cell can be improved to lower the peak surface magnetic field at the expense of a higher peak surface electric field by making the cell reentrant. Such a single-cell cavity was designed and tested at Cornell, setting a world record accelerating gradient [V. Shemelin et al., An optimized shape cavity for TESLA: concept and fabrication, 11th Workshop on RF Superconductivity, Travemuende, Germany, September 8-12, 2003; R. Geng, H. Padamsee, Reentrant cavity and first test result, Pushing the Limits of RF Superconductivity Workshop, Argonne National Laboratory, September 22-24, 2004]. However, the disadvantage to a cavity is that liquids become trapped in the reentrant portion when it is vertically hung during high pressure rinsing. While this was overcome for Cornell's single-cell cavity by flipping it several times between high pressure rinse cycles, this may not be feasible for a multi-cell cavity. One solution to this problem is to make the cavity reentrant on only one side, leaving the opposite wall angle at six degrees for fluid drainage. This idea was first presented in 2004 [T.L. Grimm et al., IEEE Transactions on Applied Superconductivity 15(6) (2005) 2393]. Preliminary designs of two new half-reentrant (HR) inner cells have since been completed, one at a high cell-to-cell coupling of 2.1% (high-k cc HR) and the other at 1.5% (low-k cc HR). The parameters of a HR cavity are comparable to a fully reentrant cavity, with the added benefit that a HR cavity can be easily cleaned with current technology

Calculation of anchor forces on penetration liners for the reactor vessel Schmehausen (Germany)

International Nuclear Information System (INIS)

Roennert, J.K.

1976-01-01

Penetrations through the walls of the single cavity PCPV Prestressed Concrete Pressure Vessel for the 300 MW(e) reactor are lined with steel penetration liners welded to the liner of the cavity. For gas-tightness of the system the penetrations are closed by covers. To secure their integration with the concrete, the liners are anchored to it by means of shear studs and/or angles. Being embedded in concrete, over the full width of the walls, the liners are exposed to lateral and longitudinal concrete deformations causing forces on the anchors. The axial blow-out force due to the pressure of the coolant on the closures must also be transferred through the anchors to the concrete. In the design of anchored penetration liners stress analyses are performed to determine anchor forces under different loading conditions and at several ages of the PCPV. The present paper deals with the mathematical estimation of the anchor forces on the basis of given concrete deformations, temperature of liners, and pressure in the vessel by the method of replacing the penetration liners and their anchors by a spring model with linear stiffness characteristics for both the liner and the anchors. An example of the computations on a digital computer is shown. (author)

In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

Energy Technology Data Exchange (ETDEWEB)

Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

2015-10-15

Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

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 porous medium model for predicting the duct wall temperature of sodium fast reactor fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Yu, Yiqi, E-mail: yyu@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Merzari, Elia; Obabko, Aleksandr [Mathematics and Computer Science Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Thomas, Justin [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States)

2015-12-15

Highlights: • The proposed models are 400 times less computationally expensive than CFD simulations. • The proposed models show good duct wall temperature agreement with CFD simulations. • The paper provides an efficient tool for coupled radial core expansion calculation. - Abstract: Porous medium models have been established for predicting duct wall temperature of sodium fast reactor rod bundle assembly, which is much less computationally expensive than conventional CFD simulations that explicitly represent the wire-wrap and fuel pin geometry. Three porous medium models are proposed in this paper. Porous medium model 1 takes the whole assembly as one porous medium of uniform characteristics in the conventional approach. Porous medium model 2 distinguishes the pins along the assembly's edge from those in the interior with two distinct regions, each with a distinct porosity, resistance, and volumetric heat source. This accounts for the different fuel-to-coolant volume ratio in the two regions, which is important for predicting the temperature of the assembly's exterior duct wall. In Porous medium model 3, a precise resistance distribution was employed to define the characteristic of the porous medium. The results show that both porous medium model 2 and 3 can capture the average duct wall temperature well. Furthermore, the local duct wall variations due to different sub-channel patterns in bare rod bundles are well captured by porous medium model 3, although the wire effect on the duct wall temperature in wire wrap rod bundle has not been fully reproduced yet.

Thermal effect of periodical bakeout on tritium inventory in first wall and permeation to coolant in reactor life

International Nuclear Information System (INIS)

Nakahara, Katsuhiko

1989-01-01

In view of safety, it is very important to control the tritium inventory in first walls and permeation to the coolant. A time-dependent diffusion and temperature calculation code, TPERM, was developed. Using this code, a numerical study on the long term effects of the bakeout temperature on tritium inventory and tritium permeation to the coolant was made. In this study, an FER type first wall (stainless steel) was considered and a cyclic operation (one cycle includes a plasma burn phase and a bakeout phase) was assumed. The results are as follows: (i) There is almost no difference in the tritium inventory in the first wall between the operation with 150 0 C-bakeout and the continuous burning operation (without bakeout). In both cases there is not tritium permeation to the coolant at 5 years' integrated burn time. The 150 0 C-bakeout is effective to release tritium in the surface (to 0.1 mm depth) region on the plasma side, but it is not effective to decrease the tritium inventory over the reactor life. (ii) To decrease the tritium inventory, a bakeout at a temperature higher than 150 0 C is necessary. But a high temperature bakeout causes earlier tritium permeation to the coolant. (iii) From these results it is suggested that the decrease the tritium inventory over the reactor life by bakeout, some form of protection against tritium permeation or a decontamination device in the cooling (or bakeout) system becomes necessary. (orig.)

Cavity formation by the impact of Leidenfrost spheres

KAUST Repository

Marston, Jeremy

2012-05-01

We report observations of cavity formation and subsequent collapse when a heated sphere impacts onto a liquid pool. When the sphere temperature is much greater than the boiling point of the liquid, we observe an inverted Leidenfrost effect where the sphere is encompassed by a vapour layer that prevents physical contact with the liquid. This creates the ultimate non-wetting scenario during sphere penetration through a free surface, producing very smooth cavity walls. In some cases during initial entry, however, the liquid contacts the sphere at the equator, leading to the formation of a dual cavity structure. For cold sphere impacts, where a contact line is observed, we reveal details of the contact line pinning, which initially forms a sawtooth pattern. We also observe surface waves on the cavity interface for cold spheres. We compare our experimental results to previous studies of cavity dynamics and, in particular, the influence of hydrophobicity on the entry of the sphere. © 2012 Cambridge University Press.

Nuclear reactor installation

International Nuclear Information System (INIS)

Jungmann, A.

1976-01-01

A nuclear reactor metal pressure vessel is surrounded by a concrete wall forming an annular space around the vessel. Thermal insulation is in this space and surrounds the vessel, and a coolant-conductive layer is also in this space surrounding the thermal insulation, coolant forced through this layer reducing the thermal stress on the concrete wall. The coolant-conductive layer is formed by concrete blocks laid together and having coolant passages, these blocks being small enough individually to permit them to be cast from concrete at the reactor installation, the thermal insulation being formed by much larger sheet-metal clad concrete segments. Mortar is injected between the interfaces of the coolant-conductive layer and concrete wall and the interfaces between the fluid-conductive layer and the insulation, a layer of slippery sheet material being interposed between the insulation and the mortar. When the pressure vessel is thermally expanded by reactor operation, the annular space between it and the concrete wall is completely filled by these components so that zero-excursion rupture safeguard is provided for the vessel. 4 claims, 1 figure

Accoustic Localization of Breakdown in Radio Frequency Accelerating Cavities

Energy Technology Data Exchange (ETDEWEB)

Lane, Peter Gwin [IIT, Chicago

2016-07-01

Current designs for muon accelerators require high-gradient radio frequency (RF) cavities to be placed in solenoidal magnetic fields. These fields help contain and efficiently reduce the phase space volume of source muons in order to create a usable muon beam for collider and neutrino experiments. In this context and in general, the use of RF cavities in strong magnetic fields has its challenges. It has been found that placing normal conducting RF cavities in strong magnetic fields reduces the threshold at which RF cavity breakdown occurs. To aid the effort to study RF cavity breakdown in magnetic fields, it would be helpful to have a diagnostic tool which can localize the source of breakdown sparks inside the cavity. These sparks generate thermal shocks to small regions of the inner cavity wall that can be detected and localized using microphones attached to the outer cavity surface. Details on RF cavity sound sources as well as the hardware, software, and algorithms used to localize the source of sound emitted from breakdown thermal shocks are presented. In addition, results from simulations and experiments on three RF cavities, namely the Aluminum Mock Cavity, the High-Pressure Cavity, and the Modular Cavity, are also given. These results demonstrate the validity and effectiveness of the described technique for acoustic localization of breakdown.

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.)

RF Power Requirements for PEFP SRF Cavity Test

International Nuclear Information System (INIS)

Kim, Han Sung; Seol, Kyung Tae; Kwon, Hyeok Jung; Cho, Yong Sub

2011-01-01

For the future extension of the PEFP (Proton Engineering Frontier Project) Proton linac, preliminary study on the SRF (superconducting radio-frequency) cavity is going on including a five-cell prototype cavity development to confirm the design and fabrication procedures and to check the RF and mechanical properties of a low-beta elliptical cavity. The main parameters of the cavity are like followings. - Frequency: 700 MHz - Operating mode: TM010 pi mode - Cavity type: Elliptical - Geometrical beta: 0.42 - Number of cells: 5 - Accelerating gradient: 8 MV/m - Epeak/Eacc: 3.71 - Bpeak/Eacc: 7.47 mT/(MV/m) - R/Q: 102.3 ohm - Epeak: 29.68 MV/m (1.21 Kilp.) - Geometrical factor: 121.68 ohm - Cavity wall thickness: 4.3 mm - Stiffening structure: Double ring - Effective length: 0.45 m For the test of the cavity at low temperature of 4.2 K, many subsystems are required such as a cryogenic system, RF system, vacuum system and radiation shielding. RF power required to generate accelerating field inside cavity depends on the RF coupling parameters of the power coupler and quality factor of the SRF cavity and the quality factor itself is affected by several factors such as operating temperature, external magnetic field level and surface condition. Therefore, these factors should be considered to estimate the required RF power for the SRF cavity test

Moisture Management for High R-Value Walls

Energy Technology Data Exchange (ETDEWEB)

Lepage, R.; Schumacher, C.; Lukachko, A.

2013-11-01

The following report explains the moisture-related concerns for High R-value wall assemblies and discusses past Building America research work that informs this study. Hygrothermal simulations were prepared for several common approaches to High R-value wall construction in six cities (Houston, Atlanta, Seattle, St. Louis, Chicago, and International Falls) representing a range of climate zones (2, 3, 4C, 4, 5A, and 7, respectively). The simulations are informed by experience gained from past research in this area and validated by field measurement and forensic experience. The modeling program was developed to assess the moisture durability of the wall assemblies based on three primary sources of moisture: construction moisture, air leakage condensation, and bulk water leakage. The peak annual moisture content of the wood based exterior sheathing was used to comparatively analyze the response to the moisture loads for each of the walls in each given city. Walls which experienced sheathing moisture contents between 20% and 28% were identified as risky, whereas those exceeding 28% were identified as very high risk. All of the wall assemblies perform well under idealized conditions. However, only the walls with exterior insulation, or cavity insulation which provides a hygrothermal function similar to exterior insulation, perform adequately when exposed to moisture loads. Walls with only cavity insulation are particularly susceptible to air leakage condensation. None of the walls performed well when a precipitation based bulk water leak was introduced to the backside of the sheathing, emphasizing the importance of proper flashing details.

Hydroforming of elliptical cavities

Science.gov (United States)

Singer, W.; Singer, X.; Jelezov, I.; Kneisel, P.

2015-02-01

fabricated. The clad seamless tubes were produced using hot bonding or explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 and 3 mm respectively. The rf performance of the best NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40 MV /m . The advantages and disadvantages of hydroformed cavities are discussed in this paper.

Superconducting cavity development at RRCAT

International Nuclear Information System (INIS)

Joshi, S.C.

2015-01-01

Raja Ramanna Centre for Advanced Technology (RRCAT), Indore pursuing a program on 'R and D Activities for High Energy Proton Linac based Spallation Neutron Source'. Spallation neutron source (SNS) facility will provide high flux pulse neutrons for research in the areas of condensed matter physics, materials science, chemistry, biology and engineering. This will complement the existing synchrotron light source facility, INDUS-2 at RRCAT and reactor based neutron facilities at BARC. RRCAT is also participating in approved mega project on 'Physics and Advanced Technology for High Intensity Proton Accelerator' to support activities of Indian Institutions - Fermilab Collaboration (IIFC). The SNS facility will have a 1 GeV superconducting proton injector linac and 1 GeV accumulator ring. The linac will comprise of large number of superconducting radio-frequency (SCRF) cavities operating at different RF frequencies housed in suitable cryomodules. Thus, an extensive SCRF cavity infrastructure setup is being established. In addition, a scientific and technical expertise are also being developed for fabrication, processing and testing of the SCRF cavities for series production. The paper presents the status of superconducting cavity development at RRCAT

Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

International Nuclear Information System (INIS)

Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

2004-01-01

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

Use of the upper radial order modes in spherical superconducting cavities

International Nuclear Information System (INIS)

Reuss, J.

1975-04-01

Spherical cavities resonating on a high g radial order mode are considered. The ratio of the maximum magnetic field inside the cavity to the maximum field on the wall is proportional to g. The proportion coefficient is given for the TEsub(g10); TEsub(g20), TMsub(g10), and TMsub(g20) modes. That corresponds to an energy concentration at the center. Owing to this property the superconducting cavities might be used to produce strong H.F. magnetic fields (larger than 10 Teslas) [fr

Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

International Nuclear Information System (INIS)

Chichindaev, D.A.

2001-01-01

An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

PEP-II RF Cavity Revisited (LCC-0032)

Energy Technology Data Exchange (ETDEWEB)

Rimmer, R.

2004-03-23

This report describes the results of numerical simulations of the PEP-II RF cavity performed after the completion of the construction phase of the project and comparisons are made to previous calculations and measured results. These analyses were performed to evaluate new calculation techniques for the HOM distribution and RF surface heating that were not available at the time of the original design. These include the use of a high frequency electromagnetic element in ANSYS and the new Omega 3P code to study wall losses, and the development of broadband time domain simulation methods in MAFIA for the HOM loading. The computed HOM spectrum is compared with cavity measurements and observed beam-induced signals. The cavity fabrication method is reviewed, with the benefit of hindsight, and simplifications are discussed.

Modeling the Rapid Boil-Off of a Cryogenic Liquid When Injected into a Low Pressure Cavity

Science.gov (United States)

Lira, Eric

2016-01-01

Many launch vehicle cryogenic applications require the modeling of injecting a cryogenic liquid into a low pressure cavity. The difficulty of such analyses lies in accurately predicting the heat transfer coefficient between the cold liquid and a warm wall in a low pressure environment. The heat transfer coefficient and the behavior of the liquid is highly dependent on the mass flow rate into the cavity, the cavity wall temperature and the cavity volume. Testing was performed to correlate the modeling performed using Thermal Desktop and Sinda Fluint Thermal and Fluids Analysis Software. This presentation shall describe a methodology to model the cryogenic process using Sinda Fluint, a description of the cryogenic test set up, a description of the test procedure and how the model was correlated to match the test results.

Molecular beam mass spectrometer equipped with a catalytic wall reactor for in situ studies in high temperature catalysis research

International Nuclear Information System (INIS)

Horn, R.; Ihmann, K.; Ihmann, J.; Jentoft, F.C.; Geske, M.; Taha, A.; Pelzer, K.; Schloegl, R.

2006-01-01

A newly developed apparatus combining a molecular beam mass spectrometer and a catalytic wall reactor is described. The setup has been developed for in situ studies of high temperature catalytic reactions (>1000 deg. C), which involve besides surface reactions also gas phase reactions in their mechanism. The goal is to identify gas phase radicals by threshold ionization. A tubular reactor, made from the catalytic material, is positioned in a vacuum chamber. Expansion of the gas through a 100 μm sampling orifice in the reactor wall into differentially pumped nozzle, skimmer, and collimator chambers leads to the formation of a molecular beam. A quadrupole mass spectrometer with electron impact ion source designed for molecular beam inlet and threshold ionization measurements is used as the analyzer. The sampling time from nozzle to detector is estimated to be less than 10 ms. A detection time resolution of up to 20 ms can be reached. The temperature of the reactor is measured by pyrometry. Besides a detailed description of the setup components and the physical background of the method, this article presents measurements showing the performance of the apparatus. After deriving the shape and width of the energy spread of the ionizing electrons from measurements on N 2 and He we estimated the detection limit in threshold ionization measurements using binary mixtures of CO in N 2 to be in the range of several hundreds of ppm. Mass spectra and threshold ionization measurements recorded during catalytic partial oxidation of methane at 1250 deg. C on a Pt catalyst are presented. The detection of CH 3 · radicals is successfully demonstrated

Molecular beam mass spectrometer equipped with a catalytic wall reactor for in situ studies in high temperature catalysis research

Science.gov (United States)

Horn, R.; Ihmann, K.; Ihmann, J.; Jentoft, F. C.; Geske, M.; Taha, A.; Pelzer, K.; Schlögl, R.

2006-05-01

A newly developed apparatus combining a molecular beam mass spectrometer and a catalytic wall reactor is described. The setup has been developed for in situ studies of high temperature catalytic reactions (>1000°C), which involve besides surface reactions also gas phase reactions in their mechanism. The goal is to identify gas phase radicals by threshold ionization. A tubular reactor, made from the catalytic material, is positioned in a vacuum chamber. Expansion of the gas through a 100μm sampling orifice in the reactor wall into differentially pumped nozzle, skimmer, and collimator chambers leads to the formation of a molecular beam. A quadrupole mass spectrometer with electron impact ion source designed for molecular beam inlet and threshold ionization measurements is used as the analyzer. The sampling time from nozzle to detector is estimated to be less than 10ms. A detection time resolution of up to 20ms can be reached. The temperature of the reactor is measured by pyrometry. Besides a detailed description of the setup components and the physical background of the method, this article presents measurements showing the performance of the apparatus. After deriving the shape and width of the energy spread of the ionizing electrons from measurements on N2 and He we estimated the detection limit in threshold ionization measurements using binary mixtures of CO in N2 to be in the range of several hundreds of ppm. Mass spectra and threshold ionization measurements recorded during catalytic partial oxidation of methane at 1250°C on a Pt catalyst are presented. The detection of CH3• radicals is successfully demonstrated.

A rugby-shaped cavity for the LMJ

International Nuclear Information System (INIS)

Vandenboomgaerde, M.; Bastian, J.; Casner, A.; Galmiche, D.; Jadaud, J.P.; Lafitte, S.; Liberatore, S.; Malinie, G.; Philippe, F.

2008-01-01

Numerical studies show that a rugby-shaped hohlraum for indirect drive laser ignition has some advantages: it allows a better symmetry for the X-ray irradiation of the central target and it required less laser power. Rugby-shaped cavities have been tested successfully at the Omega facility. The energetic advantage is all the more important as the cavity is bigger. Simulations have shown that a rugby-shaped hohlraum plus adequate materials for the intern wall plus an optimization of the central target could open the way to an ignition with only 160 laser beams at the LMJ (Megajoule Laser) facility. (A.C.)

Latest developments in prestressed concrete vessels for gas-cooled reactors

International Nuclear Information System (INIS)

Ople, F.S. Jr.

1979-01-01

This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

Cryogenic rf test of the first SRF cavity etched in an rf Ar/Cl2 plasma

Science.gov (United States)

Upadhyay, J.; Palczewski, A.; Popović, S.; Valente-Feliciano, A.-M.; Im, Do; Phillips, H. L.; Vušković, L.

2017-12-01

An apparatus and a method for etching of the inner surfaces of superconducting radio frequency (SRF) accelerator cavities are described. The apparatus is based on the reactive ion etching performed in an Ar/Cl2 cylindrical capacitive discharge with reversed asymmetry. To test the effect of the plasma etching on the cavity rf performance, a 1497 MHz single cell SRF cavity was used. The single cell cavity was mechanically polished and buffer chemically etched and then rf tested at cryogenic temperatures to provide a baseline characterization. The cavity's inner wall was then exposed to the capacitive discharge in a mixture of Argon and Chlorine. The inner wall acted as the grounded electrode, while kept at elevated temperature. The processing was accomplished by axially moving the dc-biased, corrugated inner electrode and the gas flow inlet in a step-wise manner to establish a sequence of longitudinally segmented discharges. The cavity was then tested in a standard vertical test stand at cryogenic temperatures. The rf tests and surface condition results, including the electron field emission elimination, are presented.

Nanostructural features degrading the performance of superconducting radio frequency niobium cavities revealed by TEM and EELS

OpenAIRE

Trenikhina, Y.; Romanenko, A.; Kwon, J.; Zuo, J. -M.; Zasadzinski, J. F.

2015-01-01

Nanoscale defect structure within the magnetic penetration depth of ~100nm is key to the performance limitations of niobium superconducting radio frequency (SRF) cavities. Using a unique combination of advanced thermometry during cavity RF measurements, and TEM structural and compositional characterization of the samples extracted from cavity walls, we discover the existence of nanoscale hydrides in electropolished cavities limited by the high field Q slope, and show the decreased hydride for...

Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

International Nuclear Information System (INIS)

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

1994-04-01

A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

Energy Technology Data Exchange (ETDEWEB)

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

1994-04-01

A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

Mechanical Design of a New Injector Cryomodule 2-Cell Cavity at CEBAF

Energy Technology Data Exchange (ETDEWEB)

Cheng, Guangfeng G. [JLAB; Henry, James E. [JLAB; Mammosser, John D. [JLAB; Rimmer, Robert A. [JLAB; Wang, Haipeng [JLAB; Wiseman, Mark A. [JLAB; Yang, Shuo [JLAB

2013-12-01

As a part of Jefferson Lab’s 12 GeV upgrade, a new injector superconducting RF cryomodule is required. This unit consists of a 2-cell and 7-cell cavity, with the latter being refurbished from an existing cavity. The new 2-cell cavity requires electromagnetic design and optimization followed by mechanical design analyses. The electromagnetic design is reported elsewhere. This paper aims to present the procedures and conclusions of the analyses on cavity tuning sensitivity, pressure sensitivity, upset condition pressure induced stresses, and structural vibration frequencies. The purposes of such analyses include: 1) provide reference data for cavity tuner design; 2) examine the structural integrity of the cavity; and 3) evaluate the 2-cell cavity’s resistance to microphonics. Design issues such as the location of stiffening rings, effect of tuner stiffness on cavity stress, choice of cavity wall thickness, etc. are investigated by conducting extensive finite element analyses. Progress in fabrication of the 2-cell cavity is also reported.

Energy deposition in STARFIRE reactor components

International Nuclear Information System (INIS)

Gohar, Y.; Brooks, J.N.

1985-04-01

The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry

Lifetime estimates of a fusion reactor first wall by linear damage summation and strain range partitioning methods

International Nuclear Information System (INIS)

Liu, K.C.; Grossbeck, M.L.

1979-01-01

A generalized model of a first wall made of 20% cold-worked steel was examined for neutron wall loadings ranging from 2 to 5 MW/m 2 . A spectrum of simplified on-off duty cycles was assumed with a 95% burn time. Independent evaluations of cyclic lifetimes were based on two methods: the method of linear damage summation currently being employed for use in ASME high-temperature design Code Case N-47 and that of strain range partitioning being studied for inclusion in the design code. An important point is that the latter method can incorporate a known decrease in ductility for materials subject to irradiation as a parameter, so low-cycle fatigue behavior can be estimated for irradiated material. Lifetimes predicted by the two methods agree reasonably well despite their diversity in concept. Lack of high-cycle fatigue data for the material tested at temperatures within the range of our interest precludes making conclusions on the accuracy of the predicted results, but such data are forthcoming. The analysis includes stress relaxation due to thermal and irradiation-induced creep. Reduced ductility values from irradiations that simulate the environment of the first wall of a fusion reactor were used to estimate the lifetime of the first wall under irradiation. These results indicate that 20% cold-worked type 316 stainless steel could be used as a first-wall material meeting a 8 to 10 MW-year/m 2 lifetime goal for a neutron wall loading of about 2 MW-year/m 2 and a maximum temperature of about 500 0 C

Design of a high-temperature first wall/blanket for a d-d compact Reversed-Field-Pinch reactor (CRFPR)

International Nuclear Information System (INIS)

Dabiri, A.E.; Glancy, J.E.

1983-05-01

A high-temperature first wall/blanket which would take full advantage of the absence of tritium breeding in a d-d reactor was designed. This design which produces steam at p = 7 MPa and T = 538 0 C at the blanket exit eliminates the requirement for a separate steam generator. A steam cycle with steam-to-steam reheat yielding about 37.5 percent efficiency is compatible with this design

Hydroforming of elliptical cavities

Directory of Open Access Journals (Sweden)

W. Singer

2015-02-01

double-cell cavities of the TESLA shape have been fabricated. The clad seamless tubes were produced using hot bonding or explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 and 3 mm respectively. The rf performance of the best NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40  MV/m. The advantages and disadvantages of hydroformed cavities are discussed in this paper.

Systems and methods for enhancing isolation of high-temperature reactor containments

Energy Technology Data Exchange (ETDEWEB)

Peterson, Per F.

2017-09-26

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

Fusion reactor safety studies, FY 1977

International Nuclear Information System (INIS)

Darby, J.B. Jr.

1978-04-01

This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

2017-04-01

Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

Development of wall ranging radiation inspection robot

International Nuclear Information System (INIS)

Lee, B. J.; Yoon, J. S.; Park, Y. S.; Hong, D. H.; Oh, S. C.; Jung, J. H.; Chae, K. S.

1999-03-01

With the aging of nation's nuclear facilities, the target of this project is to develop an under water wall ranging robotic vehicle which inspects the contamination level of the research reactor (TRIGA MARK III) as a preliminary process to dismantling. The developed vehicle is driven by five thrusters and consists of small sized control boards, and absolute position detector, and a radiation detector. Also, the algorithm for autonomous navigation is developed and its performance is tested through under water experiments. Also, the test result at the research reactor shows that the vehicle firmly attached the wall while measuring the contamination level of the wall

Development of wall ranging radiation inspection robot

Energy Technology Data Exchange (ETDEWEB)

Lee, B. J.; Yoon, J. S.; Park, Y. S.; Hong, D. H.; Oh, S. C.; Jung, J. H.; Chae, K. S

1999-03-01

With the aging of nation's nuclear facilities, the target of this project is to develop an under water wall ranging robotic vehicle which inspects the contamination level of the research reactor (TRIGA MARK III) as a preliminary process to dismantling. The developed vehicle is driven by five thrusters and consists of small sized control boards, and absolute position detector, and a radiation detector. Also, the algorithm for autonomous navigation is developed and its performance is tested through under water experiments. Also, the test result at the research reactor shows that the vehicle firmly attached the wall while measuring the contamination level of the wall.

Hydroforming of superconducting TESLA cavities

International Nuclear Information System (INIS)

Singer, W.; Kaiser, H.; Singer, X.

2003-01-01

Seamless fabrication of single-cell and multi-cell TESLA shape cavities by hydroforming has been developed at DESY. The forming takes place by expanding the seamless tube with internal water pressure while simultaneously swaging it axially. Tube radius and axial displacement are being computer controlled in accordance with results of FEM simulations and the experimentally obtained strain-stress curve of tube material. Several Nb single cell cavities have been produced. A first bulk Nb double cell cavity has been fabricated. The Nb seamless tubes have been produced by spinning and deep drawing. Surface treatment such as buffered chemical polishing, (BCP), electropolishing (EP), high pressure ultra pure water rinsing (HPR), annealing at 800degC and baking at ca. 150degC have been applied. The best single cell bulk Nb cavity has reached an accelerating gradient of Eacc > 42 MV/m after ca. 250 μm BCP and 100 μm EP. Several bimetallic NbCu single cell cavities of TESLA shape have been fabricated. The seamless tubes have been produced by explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 mm and 3 mm respectively. The RF performance of NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40 MV/m after ca. 180 μm BCP, annealing at 800degC and baking at 140degC for 30 hours. The degradation of the quality factor Qo after repeated quenching is moderate, after ca. 150 quenches it reaches the saturation point of Qo=1.4x10 10 at low field. This indicates that on the basis of RF performance and material costs the combination of hydroforming with tube cladding is a very promising option. (author)

DEMOS PLUS. Robot for decontaminating soils and cavity walls of the reactor and fuel pools NPP primarily during periods of recharging fuel

International Nuclear Information System (INIS)

Lacalle Bayo, J.; Vaquer Perez, J. I.; Rosello Garcia, J. I.

2014-01-01

In this work the robot Plus Demos, equipment that has been developed by GD Energy Services from the redesign and development of robot Demos show, which took place on last year. This evolution has given the team greater capabilities, highlighting the decontamination of vertical surfaces. The main objective pursued is to minimize operational doses to workers operating in cavity as well as the risk of surface contamination during them. (Author)

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Rf transfer in the Coupled-Cavity Free-Electron Laser Two-Beam Accelerator

International Nuclear Information System (INIS)

Makowski, M.A.

1991-01-01

A significant technical problem associated with the Coupled-Cavity Free-Electron Laser Two-Beam Accelerator is the transfer of RF energy from the drive accelerator to the high-gradient accelerator. Several concepts have been advanced to solve this problem. This paper examines one possible solution in which the drive and high-gradient cavities are directly coupled to one another by means of holes in the cavity walls or coupled indirectly through a third intermediate transfer cavity. Energy cascades through the cavities on a beat frequency time scale which must be made small compared to the cavity skin time but large compared to the FEL pulse length. The transfer is complicated by the fact that each of the cavities in the system can support many resonant modes near the chosen frequency of operation. A generalized set of coupled-cavity equations has been developed to model the energy transfer between the various modes in each of the cavities. For a two cavity case transfer efficiencies in excess of 95% can be achieved. 3 refs., 2 figs

Current Status on the Development of a Double Wall Tube Steam Generator

Energy Technology Data Exchange (ETDEWEB)

Nam, Ho Yun; Choi, Byoung Hae; Kim, Jong Man; Kim, Byung Ho

2007-12-15

A fast reactor, which uses sodium as a coolant, has a lot of merits as a next generation nuclear reactor. However, the possibility of a sodium-water reaction occurrence hinders the commercialization of this reactor. As one way to improve the reliability of a steam generator, a double-wall tube steam generator is being developed in GEN-4 program. In this report, the current state of the technical developments for a double-wall tube steam generator are reviewed and a future plan for the development of a double-wall tube steam generator is established. The current focuses of this research are an improvement of the heat transfer capability for a double-wall tube and the development of a proper leak detection method for the failure of a double-wall tube during a reactor operation. The ideal goal is an on-line leak detection of a double wall tube to prevent the sodium-water reaction. However, such a method is not developed as yet. An alternative method is being used to improve the reliability of a steam generator by performing a non-destructive test of a double wall tube during the refueling period of a reactor. In this method a straight double wall tube is employed to perform this test easily, but has a difficulty regarding an absorption of a thermal expansion of the used materials. If an on-line leak detection method is developed, the demerits of a straight double-wall tube are avoided by using a helical type double-wall tube, and the probability of a sodium-water reaction can be reduced to a level less than the design-based accident.

Residual stress improving method for reactor structural component and residual stress improving device therefor

Energy Technology Data Exchange (ETDEWEB)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato

1996-09-03

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Residual stress improving method for reactor structural component and residual stress improving device therefor

International Nuclear Information System (INIS)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato.

1996-01-01

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Electromagnetic characterization of superconducting radio-frequency cavities for gw detection

Science.gov (United States)

Ballantini, R.; Bernard, Ph; Chincarini, A.; Gemme, G.; Parodi, R.; Picasso, E.

2004-03-01

The electromagnetic properties of a prototype gravitational wave detector, based on two coupled superconducting microwave cavities, were tested. The radio-frequency (rf) detection system was carefully analysed. With the use of piezoelectric crystals small harmonic displacements of the cavity walls were induced and the parametric conversion of the electromagnetic field inside the cavities explored. Experimental results of bandwidth and sensitivity of the parametric converter versus stored energy and voltage applied to the piezoelectric crystal are reported. A rf control loop, developed to stabilize phase changes on signal paths, gave a 125 dBc rejection of the drive mode on a time scale of 1 h.

Electromagnetic characterization of superconducting radio-frequency cavities for gw detection

International Nuclear Information System (INIS)

Ballantini, R; Bernard, Ph; Chincarini, A; Gemme, G; Parodi, R; Picasso, E

2004-01-01

The electromagnetic properties of a prototype gravitational wave detector, based on two coupled superconducting microwave cavities, were tested. The radio-frequency (rf) detection system was carefully analysed. With the use of piezoelectric crystals small harmonic displacements of the cavity walls were induced and the parametric conversion of the electromagnetic field inside the cavities explored. Experimental results of bandwidth and sensitivity of the parametric converter versus stored energy and voltage applied to the piezoelectric crystal are reported. A rf control loop, developed to stabilize phase changes on signal paths, gave a 125 dBc rejection of the drive mode on a time scale of 1 h

Reynolds number and end-wall effects on a lid-driven cavity flow

International Nuclear Information System (INIS)

Prasad, A.K.; Koseff, J.R.

1989-01-01

A series of experiments has been conducted in a lid-driven cavity of square cross section (depth = width = 150 mm) for Reynolds numbers (Re, based on lid speed and cavity width) between 3200 and 10 000, and spanwise aspect ratios (SAR) between 0.25:1 and 1:1. Flow visualization using polystyrene beads and two-dimensional laser-Doppler anemometer (LDA) measurements have shed new light on the momentum transfer processes within the cavity. This paper focuses on the variation, with Re and SAR, of the mean and the rms velocities profiles, as well as the /similar to/(U'V') profile, along the horizontal and vertical centerlines in the symmetry plane. In addition, the contribution of the large-scale ''organized structures,'' and the high-frequency ''turbulent'' velocity fluctuations to the total rms is examined. At low Re, the organized structures account for most of the energy contained in the flow irrespective of SAR. As the Re increases, however, so does the energy content of the higher frequency fluctuations. This trend is not independent of SAR; a reduction in the SAR causes the ''organized structures'' to again become more evident

Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores

International Nuclear Information System (INIS)

El Ganaoui, K.

2006-09-01

In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)

Study of ex-vessel steam explosion risk of Reactor Pit Flooding System and structural response of containment for CPR1000"+ Unit

International Nuclear Information System (INIS)

Zhang Juanhua; Chen Peng

2015-01-01

Reactor Pit Flooding System is one of the special mitigation measures for severe accident for CPR1000"+ Unit. If the In-Vessel Relocation function of Reactor Pit Flooding System is failed, there is the steam explosion risk in reactor cavity. This paper firstly adopts MC3D code to build steam explosion model in order to calculate the pressure load and impulses of steam explosion that are as the input data of containment structural response analysis. The next step is to model the containment structure and analyze the structural response by ABAQUS code. The analysis results show that the integral damage induced by steam explosion to the external containment wall is shallow, and the containment structural integrity can be maintained. The risk and damage to the containment integrity reduced by steam explosion of RPF is small, and it does not influence the design and implementation of RPF. (author)

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

Numerical simulation of magnetic convection ferrofluid flow in a permanent magnet-inserted cavity

Science.gov (United States)

Ashouri, Majid; Behshad Shafii, Mohammad

2017-11-01

The magnetic convection heat transfer in an obstructed two-dimensional square cavity is investigated numerically. The walls of the cavity are heated with different constant temperatures at two sides, and isolated at two other sides. The cavity is filled with a high Prandtl number ferrofluid. The convective force is induced by a magnetic field gradient of a thermally insulated square permanent magnet located at the center of the cavity. The results are presented in the forms of streamlines, isotherms, and Nusselt number for various values of magnetic Rayleigh numbers and permanent magnet size. Two major circulations are generated in the cavity, clockwise flow in the upper half and counterclockwise in the lower half. In addition, strong circulations are observed around the edges of the permanent magnet surface. The strength of the circulations increase monotonically with the magnetic Rayleigh number. The circulations also increase with the permanent magnet size, but eventually, are suppressed for larger sizes. It is found that there is an optimum size for the permanent magnet due to the contrary effects of the increase in magnetic force and the increase in flow resistance by increasing the size. By increasing the magnetic Rayleigh number or isothermal walls temperature ratio, the heat transfer rate increases.

Safety equipment in a reactor

International Nuclear Information System (INIS)

Shiratori, Hirozo; Ishiyama, Satoshi; Ugawa, Yukio.

1976-01-01

Object: To safely retain, even if fuel should be molten and flown through the bottom of a container in a reactor, the molten fuel to remove heat generation of the fuel to prevent occurrence of a critical trouble. Structure: A reactor container housing a core and coolant has thereunder a separation dome in a central portion thereof and a partitioning plate coaxially and circularly disposed in the periphery of the separation dome, with a tray formed of magnesium oxide being disposed. Further, a cooling path system is provided so as to surround the tray. The cooling path system and the reactor container are surrounded and protected by a reactor wall provided with heat insulating refractory bricks, a coolant pouring system extends through the reactor wall, and the coolant is supplied to the tray. (Furukawa, Y.)

Segregation gettering by implantation-formed cavities and B-Si precipitates in silicon

International Nuclear Information System (INIS)

Myers, S.M.; Petersen, G.A.; Follstaedt, D.M.

1998-01-01

The authors show that Fe, Co, Cu, and Au in Si undergo strong segregation gettering to cavities and B-Si precipitates formed by He or B ion implantation and annealing. The respective mechanisms are argued to be chemisorption on the cavity walls and occupation of solution sites within the disordered, B-rich, B-Si phase. The strengths of the reactions are evaluated, enabling prediction of gettering performance

Near-wall serpentine cooled turbine airfoil

Science.gov (United States)

Lee, Ching-Pang

2013-09-17

A serpentine coolant flow path (54A-54G) formed by inner walls (50, 52) in a cavity (49) between pressure and suction side walls (22, 24) of a turbine airfoil (20A). A coolant flow (58) enters (56) an end of the airfoil, flows into a span-wise channel (54A), then flows forward (54B) over the inner surface of the pressure side wall, then turns behind the leading edge (26), and flows back along a forward part of the suction side wall, then follows a loop (54E) forward and back around an inner wall (52), then flows along an intermediate part of the suction side wall, then flows into an aft channel (54G) between the pressure and suction side walls, then exits the trailing edge (28). This provides cooling matched to the heating topography of the airfoil, minimizes differential thermal expansion, revives the coolant, and minimizes the flow volume needed.

Cryogenic rf test of the first SRF cavity etched in an rf Ar/Cl2 plasma

Directory of Open Access Journals (Sweden)

J. Upadhyay

2017-12-01

Full Text Available An apparatus and a method for etching of the inner surfaces of superconducting radio frequency (SRF accelerator cavities are described. The apparatus is based on the reactive ion etching performed in an Ar/Cl2 cylindrical capacitive discharge with reversed asymmetry. To test the effect of the plasma etching on the cavity rf performance, a 1497 MHz single cell SRF cavity was used. The single cell cavity was mechanically polished and buffer chemically etched and then rf tested at cryogenic temperatures to provide a baseline characterization. The cavity’s inner wall was then exposed to the capacitive discharge in a mixture of Argon and Chlorine. The inner wall acted as the grounded electrode, while kept at elevated temperature. The processing was accomplished by axially moving the dc-biased, corrugated inner electrode and the gas flow inlet in a step-wise manner to establish a sequence of longitudinally segmented discharges. The cavity was then tested in a standard vertical test stand at cryogenic temperatures. The rf tests and surface condition results, including the electron field emission elimination, are presented.

BWR type reactors

International Nuclear Information System (INIS)

Hayashi, Katsuhisa; Watanabe, Shigeru.

1983-01-01

Purpose: To simplify the structure of control rod driving systems, as well as improve the safety and maintainability thereof. Constitution: Control-rod-guide tubes are disposed vertically above the reactor core and control-rod drives are disposed further thereabove, by which the control rods are moved upwardly and downwardly from above the reactor core through the guide tubes. Further, a partitioning cylinder is provided between the inner cirumferential wall at the upper portion of a pressure vessel and the control-rod-guide tubes and a gas-liquid separator is disposed to the space between the partitioning cylinder and the pressure vessel wall, to which steams generated in the reactor core are introduced. In such a structure of the reactor, since all of the control rods are inserted or extracted by the control rod drive system from above the reactor core, if the control rod drives or the likes should fail and accidentally drop the control rods, they exert in the direction of suppressing the nuclear reaction, whereby the safety can be improved. (Sekiya, K.)

Analysis of wall-packed-bed thermal interactions

International Nuclear Information System (INIS)

Gorbis, Z.R.; Tillack, M.S.; Tehranian, F.; Abdou, M.A.

1995-01-01

One of the major issues remaining for ceramic breeder blankets involves uncertainties in heat transfer and thermomechanical interactions within the breeder and multiplier regions. Particle bed forms are considered in many reactor blanket designs for both the breeder and Be multiplier. The effective thermal conductivity of beds and the wall-bed thermal conductance are still not adequately characterized, particularly under the influence of mechanical stresses. The problem is particularly serious for the wall conductance between Be and its cladding, where the uncertainty can be greater than 50%. In this work, we describe a new model for the wall-bed conductance that treats the near-wall region as a finite-width zone. The model includes an estimate of the region porosity based on the number of contact points, and the contact area for smooth surfaces. It solves the heat conduction in a near-wall unit cell. The model is verified with existing data and used to predict the range of wall conductances expected in future simulation experiments and in reactor applications. (orig.)

Seismic behavior and design of a primary shield structure consisting of steel-plate composite (SC) walls

Energy Technology Data Exchange (ETDEWEB)

Booth, Peter N., E-mail: boothpn@purdue.edu [Lyles School of Civil Engineering, Purdue University, W. Lafayette, IN (United States); Varma, Amit H., E-mail: ahvarma@purdue.edu [Lyles School of Civil Engineering, Purdue University, W. Lafayette, IN (United States); Sener, Kadir C., E-mail: ksener@purdue.edu [Lyles School of Civil Engineering, Purdue University, W. Lafayette, IN (United States); Mori, Kentaro, E-mail: kentaro_mori@mhi.co.jp [Mitsubishi Heavy Industries, Ltd, Kobe (Japan)

2015-12-15

This paper presents an analytical evaluation of the seismic behavior and design of a unique primary shield (PSW) structure consisting of steel-plate composite (SC) walls designed for a typical pressurized water reactor (PWR) nuclear power plant. Researchers in Japan have previously conducted a reduced (1/6th) scale test of a PSW structure to evaluate its seismic (lateral) load-deformation behavior. This paper presents the development and benchmarking of a detailed 3D nonlinear inelastic finite element (NIFE) model to predict the lateral load-deformation response and behavior of the 1/6th scale test structure. The PSW structure consists of thick SC wall segments with complex and irregular geometry that surround the central reactor vessel cavity. The wall segments have three layers of steel plates (one each on the interior and exterior surfaces and one embedded in the middle) that are anchored to the concrete infill with stud anchors. The results from the 3D NIFE analyses include: (i) the lateral load-deformation behavior of the PSW structure, (ii) the progression of yielding in the steel plates, concrete cracking, formation of compression struts, and (iii) the final failure mode. These results are compared and benchmarked using experimental measurements and observations reported by Shodo et al. (2003). The analytical results provide significant insight into the lateral behavior and strength of the PSW structure, and are used for developing a design approach. This design approach starts with ACI 349 code equations for reinforced concrete shear walls and modifies them for application to the PSW structure. A simplified 3D linear elastic finite element (LEFE) model of the PSW structure is also proposed as a conventional structural analysis tool for estimating the design force demands for various load combinations.

Seismic behavior and design of a primary shield structure consisting of steel-plate composite (SC) walls

International Nuclear Information System (INIS)

Booth, Peter N.; Varma, Amit H.; Sener, Kadir C.; Mori, Kentaro

2015-01-01

This paper presents an analytical evaluation of the seismic behavior and design of a unique primary shield (PSW) structure consisting of steel-plate composite (SC) walls designed for a typical pressurized water reactor (PWR) nuclear power plant. Researchers in Japan have previously conducted a reduced (1/6th) scale test of a PSW structure to evaluate its seismic (lateral) load-deformation behavior. This paper presents the development and benchmarking of a detailed 3D nonlinear inelastic finite element (NIFE) model to predict the lateral load-deformation response and behavior of the 1/6th scale test structure. The PSW structure consists of thick SC wall segments with complex and irregular geometry that surround the central reactor vessel cavity. The wall segments have three layers of steel plates (one each on the interior and exterior surfaces and one embedded in the middle) that are anchored to the concrete infill with stud anchors. The results from the 3D NIFE analyses include: (i) the lateral load-deformation behavior of the PSW structure, (ii) the progression of yielding in the steel plates, concrete cracking, formation of compression struts, and (iii) the final failure mode. These results are compared and benchmarked using experimental measurements and observations reported by Shodo et al. (2003). The analytical results provide significant insight into the lateral behavior and strength of the PSW structure, and are used for developing a design approach. This design approach starts with ACI 349 code equations for reinforced concrete shear walls and modifies them for application to the PSW structure. A simplified 3D linear elastic finite element (LEFE) model of the PSW structure is also proposed as a conventional structural analysis tool for estimating the design force demands for various load combinations.

Low energy booster radio frequency cavity structural analysis

International Nuclear Information System (INIS)

Jones, K.

1994-01-01

The structural design of the Superconducting Super Collider Low Energy Booster (LEB) Radio Frequency (RF) Cavity is very unique. The cavity is made of three different materials which all contribute to its structural strength while at the same time providing a good medium for magnetic properties. Its outer conductor is made of thin walled stainless steel which is later copper plated to reduce the electrical losses. Its tuner housing is made of a fiber reinforced composite laminate, similar to G10, glued to stainless steel plating. The stainless steel of the tuner is slotted to significantly diminish the magnetically-induced eddy currents. The composite laminate is bonded to the stainless steel to restore the structural strength that was lost in slotting. The composite laminate is also a barrier against leakage of the pressurized internal ferrite coolant fluid. The cavity's inner conductor, made of copper and stainless steel, is subjected to high heat loads and must be liquid cooled. The requirements of the Cavity are very stringent and driven primarily by deflection, natural frequency and temperature. Therefore, very intricate finite element analysis was used to complement conventional hand analysis in the design of the cavity. Structural testing of the assembled prototype cavity is planned to demonstrate the compliance of the cavity design to all of its requirements

Low energy booster radio frequency cavity structural analysis

International Nuclear Information System (INIS)

Jones, K.

1993-04-01

The structural design of the Superconducting Super Collider Low Energy Booster (LEB) Radio Frequency (RF) Cavity is very unique. The cavity is made of three different materials which all contribute to its structural strength while at the same time providing a good medium for magnetic properties. Its outer conductor is made of thin walled stainless steel which is later copper plated to reduce the electrical losses. Its tuner housing is made of a fiber reinforced composite laminate, similar to G10, glued to stainless steel plating. The stainless steel of the tuner is slotted to significantly diminish the magnetically-induced eddy currents. The composite laminate is bonded to the stainless steel to restore the structural strength that was lost in slotting. The composite laminate is also a barrier against leakage of the pressurized internal ferrite coolant fluid. The cavity's inner conductor, made of copper and stainless steel, is subjected to high heat loads and must be liquid cooled. The requirements of the Cavity are very stringent and driven primarily by deflection, natural frequency and temperature. Therefore, very intricate finite element analysis was used to complement conventional hand analysis in the design of the cavity. Structural testing of the assembled prototype cavity is planned to demonstrate the compliance of the cavity design to all of its requirements

Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, Winfried

1997-01-01

This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

Multifunctional carbon nanotubes with nanoparticles embedded in their walls

International Nuclear Information System (INIS)

Mattia, D; Korneva, G; Sabur, A; Friedman, G; Gogotsi, Y

2007-01-01

Controlled amounts of nanoparticles ranging in size and composition were embedded in the walls of carbon nanotubes during a template-assisted chemical vapour deposition (CVD) process. The encapsulation of gold nanoparticles enabled surface enhanced Raman spectroscopy (SERS) detection of glycine inside the cavity of the nanotubes. Iron oxide particles are partially reduced to metallic iron during the CVD process giving the nanotubes ferromagnetic behaviour. At high nanoparticle concentrations, particle agglomerates can form. These agglomerates or larger particles, which are only partially embedded in the walls of the nanotubes, are covered by additional carbon layers inside the hollow cavity of the tube producing hillocks inside the nanotubes, with sizes comparable to the bore of the tube

Characterization of graded iron / tungsten layers for the first wall of fusion reactors

International Nuclear Information System (INIS)

Heuer, Simon

2017-01-01

The nuclear fusion has great potential to enable a CO 2 -neutral energy supply of future generations. The technical utilization of this energy source has hitherto been a challenge. In particular, high thermal loads and neutron-induced damage lead to extreme demands on the choice of materials for plasma-facing components (PFCs). These are therefore, as currently understood, made from a tungsten protective layer which is joined to a structure of low activation ferritic-martensitic (LAFM) steel. Due to the discrete transition of material properties at the LAFM-W joining zone as well as thermal loads, macroscopic stresses and plastic strains arise here. A feasible way to reduce this is to implement an intermediate layer with graded LAFM / W ratio, a so-called functional graded material (FGM). In the present work, macro-stresses and strains in the first wall of the fusion reactor DEMO are examined and evaluated by means of a finite element simulation. In this framework model components with and without graded interlayer are taken into account and the advantage of a FGM is emphasized. Parameter studies serve as a constructive guideline for the structural implementation of FGMs and components of the first wall. In addition, the feasibility of four methods (magnetron sputtering, liquid phase infiltration, modified atmospheric plasma spraying and electrodischarge sintering) with respect to the fabrication of FGMs is being studied. The resulting layers are microstructurally, thermo-physically and mechanically examined in detail. Based on this characterization and the finite element simulation, their suitability as a graded layer in the first wall of DEMO is evaluated and finally compared with alternative joining systems that are currently being tested in the research environment. [de

Resonant-frequency discharge in a multi-cell radio frequency cavity

International Nuclear Information System (INIS)

Popović, S.; Upadhyay, J.; Nikolić, M.; Vušković, L.; Mammosser, J.

2014-01-01

We are reporting experimental results on a microwave discharge operating at resonant frequency in a multi-cell radio frequency (RF) accelerator cavity. Although the discharge operated at room temperature, the setup was constructed so that it could be used for plasma generation and processing in fully assembled active superconducting radio-frequency cryo-module. This discharge offers a mechanism for removal of a variety of contaminants, organic or oxide layers, and residual particulates from the interior surface of RF cavities through the interaction of plasma-generated radicals with the cavity walls. We describe resonant RF breakdown conditions and address the issues related to resonant detuning due to sustained multi-cell cavity plasma. We have determined breakdown conditions in the cavity, which was acting as a plasma vessel with distorted cylindrical geometry. We discuss the spectroscopic data taken during plasma removal of contaminants and use them to evaluate plasma parameters, characterize the process, and estimate the volatile contaminant product removal

Resonant-frequency discharge in a multi-cell radio frequency cavity

Energy Technology Data Exchange (ETDEWEB)

Popovic, S; Upadhyay, J; Mammosser, J; Nikolic, M; Vuskovic, L

2014-11-07

We are reporting experimental results on microwave discharge operating at resonant frequency in a multi-cell radio frequency (RF) accelerator cavity. Although the discharge operated at room temperature, the setup was constructed so that it could be used for plasma generation and processing in fully assembled active superconducting radio-frequency (SRF) cryomodule (in situ operation). This discharge offers an efficient mechanism for removal of a variety of contaminants, organic or oxide layers, and residual particulates from the interior surface of RF cavities through the interaction of plasma-generated radicals with the cavity walls. We describe resonant RF breakdown conditions and address the problems related to generation and sustaining the multi-cell cavity plasma, which are breakdown and resonant detuning. We have determined breakdown conditions in the cavity, which was acting as a plasma vessel with distorted cylindrical geometry. We discuss the spectroscopic data taken during plasma removal of contaminants and use them to evaluate plasma parameters, characterize the process, and estimate the volatile contaminant product removal.

Boundary-Layer Effects on Acoustic Transmission Through Narrow Slit Cavities.

Science.gov (United States)

Ward, G P; Lovelock, R K; Murray, A R J; Hibbins, A P; Sambles, J R; Smith, J D

2015-07-24

We explore the slit-width dependence of the resonant transmission of sound in air through both a slit array formed of aluminum slats and a single open-ended slit cavity in an aluminum plate. Our experimental results accord well with Lord Rayleigh's theory concerning how thin viscous and thermal boundary layers at a slit's walls affect the acoustic wave across the whole slit cavity. By measuring accurately the frequencies of the Fabry-Perot-like cavity resonances, we find a significant 5% reduction in the effective speed of sound through the slits when an individual viscous boundary layer occupies only 5% of the total slit width. Importantly, this effect is true for any airborne slit cavity, with the reduction being achieved despite the slit width being on a far larger scale than an individual boundary layer's thickness. This work demonstrates that the recent prevalent loss-free treatment of narrow slit cavities within acoustic metamaterials is unrealistic.

Study of Low Work Function Materials for Hot Cavity Resonance Ionization Laser Ion Sources

CERN Document Server

Catherall, R; Fedosseev, V; Marsh, B; Mattolat, C; Menna, Mariano; Österdahl, F; Raeder, S; Schwellnus, F; Stora, T; Wendt, K; CERN. Geneva. AB Department

2008-01-01

The selectivity of a hot cavity resonance ionization laser ion source (RILIS) is most often limited by contributions from competing surface ionization on the hot walls of the ionization cavity. In this article we present investigations on the properties of designated high-temperature, low-work function materials regarding their performance and suitability as cavity material for RILIS. Tungsten test cavities, impregnated with a mixture of barium oxide and strontium oxide (BaOSrO on W), or alternatively gadolinium hexaboride (GdB6) were studied in comparison to a standard tungsten RILIS cavity as being routinely used for hot cavity laser ionization at ISOLDE. Measurement campaigns took place at the off-line mass separators at ISOLDE / CERN, Geneva and RISIKO / University of Mainz.

Study of low work function materials for hot cavity resonance ionization laser ion sources

CERN Document Server

Schwellnus, F; Crepieux, B; Fedosseev, V N; Marsh, B A; Mattolat, Ch; Menna, M; Österdahl, F K; Raeder, S; Stora, T; Wendta, K

2009-01-01

The selectivity of a hot cavity resonance ionization laser ion source (RILIS) is most often limited by contributions from competing surface ionization of the hot walls of the ionization cavity. In this article we present investigations on the properties of designated high temperature, low work function materials regarding their performance and suitability as cavity material for RILIS. Tungsten test cavities, impregnated with a mixture of barium oxide and strontium oxide (BaOSrO on W), or alternatively gadolinium hexaboride (GdB6) were studied in comparison to a standard tungsten RILIS cavity as being routinely used for hot cavity laser ionization at ISOLDE. Measurement campaigns took place at the off-line mass separators at ISOLDE/CERN, Geneva and RISIKO/University of Mainz.

Field dependent surface resistance of niobium on copper cavities

Directory of Open Access Journals (Sweden)

T. Junginger

2015-07-01

Full Text Available The surface resistance R_{S} of superconducting cavities prepared by sputter coating a niobium film on a copper substrate increases significantly stronger with the applied rf field compared to cavities of bulk material. A possible cause is that the thermal boundary resistance between the copper substrate and the niobium film induces heating of the inner cavity wall, resulting in a higher R_{S}. Introducing helium gas in the cavity, and measuring its pressure as a function of applied field allowed to conclude that the inner surface of the cavity is heated up by less than 120 mK when R_{S} increases with E_{acc} by 100  nΩ. This is more than one order of magnitude less than what one would expect from global heating. Additionally, the effects of cooldown speed and low temperature baking have been investigated in the framework of these experiments. It is shown that for the current state of the art niobium on copper cavities there is only a detrimental effect of low temperature baking. A fast cooldown results in a lowered R_{S}.

Thermosyphoning analysis with the CATHENA model of the blanket and first wall cooling loop for the SEAFP reactor design

International Nuclear Information System (INIS)

Ross, W.E.

1994-02-01

This report documents the thermosyphoning analysis which was performed with the CATHENA network model of one of the blanket and first wall cooling loops of the SEAFP reactor design. This thermosyphoning analysis includes four simulations, each with a slightly different model feature or assumption. These simulations are performed to assess the primary heat transport system behaviour for a complete loss of electrical power event (total loss of flow) and to estimate the rate and extent of heat-up of the incore components. For each event, a description of some of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and outflow, and first wall and blanket temperatures is provided. (author). 4 refs., 2 tabs., 32 figs

Apparatus and method for plasma processing of SRF cavities

Science.gov (United States)

Upadhyay, J.; Im, Do; Peshl, J.; Bašović, M.; Popović, S.; Valente-Feliciano, A.-M.; Phillips, L.; Vušković, L.

2016-05-01

An apparatus and a method are described for plasma etching of the inner surface of superconducting radio frequency (SRF) cavities. Accelerator SRF cavities are formed into a variable-diameter cylindrical structure made of bulk niobium, for resonant generation of the particle accelerating field. The etch rate non-uniformity due to depletion of the radicals has been overcome by the simultaneous movement of the gas flow inlet and the inner electrode. An effective shape of the inner electrode to reduce the plasma asymmetry for the coaxial cylindrical rf plasma reactor is determined and implemented in the cavity processing method. The processing was accomplished by moving axially the inner electrode and the gas flow inlet in a step-wise way to establish segmented plasma columns. The test structure was a pillbox cavity made of steel of similar dimension to the standard SRF cavity. This was adopted to experimentally verify the plasma surface reaction on cylindrical structures with variable diameter using the segmented plasma generation approach. The pill box cavity is filled with niobium ring- and disk-type samples and the etch rate of these samples was measured.

An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

Energy Technology Data Exchange (ETDEWEB)

Haihua Zhao; Per F. Peterson

2010-10-01

Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

Analysis of the Magnetic Field Effect on Entropy Generation at Thermosolutal Convection in a Square Cavity

Directory of Open Access Journals (Sweden)

Ammar Ben Brahim

2011-05-01

Full Text Available Thermosolutal convection in a square cavity filled with air and submitted to an inclined magnetic field is investigated numerically. The cavity is heated and cooled along the active walls with a mass gradient whereas the two other walls of the cavity are adiabatic and insulated. Entropy generation due to heat and mass transfer, fluid friction and magnetic effect has been determined in transient state for laminar flow by solving numerically the continuity, momentum energy and mass balance equations, using a Control Volume Finite—Element Method. The structure of the studied flows depends on four dimensionless parameters which are the Grashof number, the buoyancy ratio, the Hartman number and the inclination angle. The results show that the magnetic field parameter has a retarding effect on the flow in the cavity and this lead to a decrease of entropy generation, Temperature and concentration decrease with increasing value of the magnetic field parameter.

Integral reactor system and method for fuel cells

Science.gov (United States)

Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

2013-11-19

A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

International Nuclear Information System (INIS)

Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

1976-05-01

The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

Shear Layer Dynamics in Resonating Cavity Flows

National Research Council Canada - National Science Library

Ukeiley, Lawrence

2004-01-01

.... The PIV data was also combined with the surface pressure measurements through the application of the Quadratic Stochastic Estimation procedure to provide time resolved snapshots of the flow field. Examination of these results indicate the strong pumping action of the cavity regardless of whether resonance existed and was used to visualize the large scale structures interacting with the aft wall.

First-wall design limitations for linear magnetic fusion (LMF) reactors

International Nuclear Information System (INIS)

Gryczkowski, G.E.; Krakowski, R.A.; Steinhauer, L.C.; Zumdieck, J.

1978-01-01

One approach to the endloss problem in linear magnetic fusion (LMF) uses high magnetic field to reduce the required confinement time. This approach is limited by magnet stresses and bremsstrahlung heating of the first wall; the first-wall thermal-pulsing issue is addressed. Pertinent thermophysical parameters are developed in the context of high-field LMF to identify promising first-wall materials, and thermal fatigue experiments relevant to LMF first walls are reviewed. High-flux first-wall concepts are described which include both solid and evaporating first-wall configurations

Symmetry-Induced Light Confinement in a Photonic Quasicrystal-Based Mirrorless Cavity

Directory of Open Access Journals (Sweden)

Gianluigi Zito

2016-09-01

Full Text Available We numerically investigate the electromagnetic field localization in a two-dimensional photonic quasicrystal generated with a holographic tiling. We demonstrate that light confinement can be induced into an air mirrorless cavity by the inherent symmetry of the spatial distribution of the dielectric scatterers forming the side walls of the open cavity. Furthermore, the propagation direction can be controlled by suitable designs of the structure. This opens up new avenues for designing photonic materials and devices.

Experimental analysis of natural convection in a cavity with relation 2:1

International Nuclear Information System (INIS)

Reyes S, M.

1994-01-01

This work develop an experimental study of the natural convection in Transient State in a cavity of the relation 2:1 (long-height), heated by a heat flux on a side wall with the opposite wall at constant temperature and equal at the temperature of the fluid. The experimental work was made for a Rayleigh number of approximately 10 9 , and the Prandtl number of 7.69. The work objective is to describe the velocity fields by mean of optic methods at different times, wide of limit layers, and searching the best visual conditions for know widely the phenomena in study. We carry out a comparison of the experimental results with the analysis of scales of Patterson and Imberger (9), with the adaptations of Poujol (19), for the condition of a constant heat flux, given this theories good results. The experimental work it have the formation of a vortex near of the hot wall, this vortex, decrease only in size during the heat transfer. In the top of the cavity in the right corner we found a divergence zone such as a H ydraulic jump , mentioned by Ivey (13), and we found too a second vortex in the bottom of the wall with constant temperature, that decrease and finally disappear when the fluid reach a permanent state. This work contribute to the mechanical design of the cavity, and at the description of the best photographic conditions for the study of the natural convection, giving good results for the study of the limit layers, thermic, hydrodynamic and the intrusion. (Author)