WorldWideScience

Sample records for reactor cavity walls

  1. Scaling of reactor cavity wall loads and stresses

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.

    1977-11-01

    Scalings of reactor cavity wall loads and stresses are determined by deriving an analytic expression in terms of relevant parameters for each loading induced in the reactor cavity walls by fuel pellet microexplosion and by deriving associated expressions relating resulting stresses to shell thicknesses. Also identified are problems that require additional investigations to obtain satisfactory explicit stress estimates for the reactor cavity walls

  2. Nuclear reactor cavity streaming shield

    International Nuclear Information System (INIS)

    Klotz, R.J.; Stephen, D.W.

    1978-01-01

    The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

  3. Improved reactor cavity

    International Nuclear Information System (INIS)

    Katz, L.R.; Demarchais, W.E.

    1984-01-01

    A reactor pressure vessel disposed in a cavity has coolant inlet or outlet pipes extending through passages in the cavity walls and welded to pressure nozzles. The cavity wall has means for directing fluid away from a break at a weld away from the pressure vessel, and means for inhibiting flow of fluid toward the vessel. (author)

  4. 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

  5. High-R Walls for Remodeling: Wall Cavity Moisture Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Wiehagen, J.; Kochkin, V.

    2012-12-01

    The focus of the study is on the performance of wall systems, and in particular, the moisture characteristics inside the wall cavity and in the wood sheathing. Furthermore, while this research will initially address new home construction, the goal is to address potential moisture issues in wall cavities of existing homes when insulation and air sealing improvements are made.

  6. High-R Walls for Remodeling. Wall Cavity Moisture Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Wiehagen, J. [NAHB Research Center Industry Partnership, Upper Marlboro, MD (United States); Kochkin, V. [NAHB Research Center Industry Partnership, Upper Marlboro, MD (United States)

    2012-12-01

    The focus of the study is on the performance of wall systems, and in particular, the moisture characteristics inside the wall cavity and in the wood sheathing. Furthermore, while this research will initially address new home construction, the goal is to address potential moisture issues in wall cavities of existing homes when insulation and air sealing improvements are made.

  7. Design study of 'HIBLIC-I' reactor cavity

    International Nuclear Information System (INIS)

    Fujiie, Y.

    1984-01-01

    A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

  8. 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.

  9. Quality labels for retrofit cavity wall insulation : a comparative analysis

    NARCIS (Netherlands)

    Rovers, Twan Johannes Hendrikus; Entrop, Alexis Gerardus; Halman, Johannes I.M.

    2017-01-01

    Retrofit cavity wall insulation can be exerted to reduce the energy use for space heating and cooling of existing buildings. In multiple countries, quality labels have emerged for this insulation service. In this research project, an evaluation framework for cavity wall insulation is developed by

  10. 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.)

  11. 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.)

  12. 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)

  13. A method for detecting fungal contaminants in wall cavities.

    Science.gov (United States)

    Spurgeon, Joe C

    2003-01-01

    This article describes a practical method for detecting the presence of both fungal spores and culturable fungi in wall cavities. Culturable fungi were collected in 25 mm cassettes containing 0.8 microm mixed cellulose ester filters using aggressive sampling conditions. Both culturable fungi and fungal spores were collected in modified slotted-disk cassettes. The sample volume was 4 L. The filters were examined microscopically and dilution plated onto multiple culture media. Collecting airborne samples in filter cassettes was an effective method for assessing wall cavities for fungal contaminants, especially because this method allowed the sample to be analyzed by both microscopy and culture media. Assessment criteria were developed that allowed the sample results to be used to classify wall cavities as either uncontaminated or contaminated. As a criterion, wall cavities with concentrations of culturable fungi below the limit of detection (LOD) were classified as uncontaminated, whereas those cavities with detectable concentrations of culturable fungi were classified as contaminated. A total of 150 wall cavities was sampled as part of a field project. The concentrations of culturable fungi were below the LOD in 34% of the samples, whereas Aspergillus and/or Penicillium were the only fungal genera detected in 69% of the samples in which culturable fungi were detected. Spore counting resulted in the detection of Stachybotrys-like spores in 25% of the samples that were analyzed, whereas Stachybotrys chartarum colonies were only detected on 2% of malt extract agar plates and on 6% of corn meal agar plates.

  14. 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

  15. Permanent seal ring for a nuclear reactor cavity

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Marshall, J.R.

    1988-01-01

    A nuclear reactor containment arrangement is described including: a. a reactor vessel which thermally expands and contracts during cyclic operation of the reactor and which has a peripheral wall; b. a containment wall spaced apart from and surrounding the peripheral wall of the reactor vessel and defining an annular thermal expansion gap therebetween for accommodating thermal expansion; and c. an annular ring seal which sealingly engages and is affixed to and extends between the peripheral wall of the reactor vessel and the containment wall

  16. A Study on the Flow Characterization in the Reactor Cavity

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jung; Ko, Kwang Jeok; Kim, Sung Hwan; Kim, Min Gyu; Cho, Yeon Ho; Kim, Hyun Min [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

    2016-10-15

    In this study, the flow characterization of the cooling air in reactor cavity nearby RCPSA has been analyzed by using a 3 dimensional model and the ANSYS CFX software in order to predict the Convective Heat Transfer Coefficient (CHTC) of the RCPSA. The Reactor Cavity is the annular space by the concrete structure, the Reactor Cavity Pool Seal Assembly (RCPSA), which consists of the welded steel and is designed to be installed between the RV and the refueling pool floor, and the Reactor Vessel (RV). For such reason, the RCPSA should be designed to provide the cooling air passage for ventilation to circulate high temperature air passing by the RV during the reactor operation. It means that the RCPSA is influenced by the convection of cooling air and the thermal expansion of the RV. Therefore, the flow characterization at the reactor cavity is one of the factors of the RCPSA design during the reactor operation. The flow distribution of the cooling air in reactor cavity nearby RCPSA has been analyzed using ANSYS CFX software to obtain the CHTC at surface of the RCPSA. 1) The temperature from the RV and the insulation is one of the critical factors for the thermal gradient of the cooling air and the CHTC in the reactor cavity. 2) The rapid change of the CHTC in inner region nearby inner and outer flexure is related to the geometry shape of the RCPSA and velocity of cooling air.

  17. Natural convection and wall radiation in tall cavities

    Energy Technology Data Exchange (ETDEWEB)

    Balaji, C [Regional Engineering College, Tiruchirapalli (India). Dept. of Mechanical Engineering; Venkateshan, S P [Indian Inst. of Tech., Madras (India). Dept. of Mechanical Engineering

    1996-12-01

    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.

  18. 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

  19. Technical evaluation of corium cooling at the reactor cavity

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Chan, Eun Sun; Lee, Jae Hun; Lee, Jong In

    1998-01-01

    To terminate the progression of the severe accident and mitigate the accident consequences, corium cooling has been suggested as one of most important design features considered in the severe accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially, the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the corium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study, technical descriptions are briefly described for the important issues related to the corium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluations for those items have been performed considering present technical levels including the experiment and analytical works

  20. 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.

  1. 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

  2. Analysis of short-term reactor cavity transient

    International Nuclear Information System (INIS)

    Cheng, T.C.; Fischer, S.R.

    1981-01-01

    Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity

  3. Effect of Axisymmetric Aft Wall Angle Cavity in Supersonic Flow Field

    Science.gov (United States)

    Jeyakumar, S.; Assis, Shan M.; Jayaraman, K.

    2018-03-01

    Cavity plays a significant role in scramjet combustors to enhance mixing and flame holding of supersonic streams. In this study, the characteristics of axisymmetric cavity with varying aft wall angles in a non-reacting supersonic flow field are experimentally investigated. The experiments are conducted in a blow-down type supersonic flow facility. The facility consists of a supersonic nozzle followed by a circular cross sectional duct. The axisymmetric cavity is incorporated inside the duct. Cavity aft wall is inclined with two consecutive angles. The performance of the aft wall cavities are compared with rectangular cavity. Decreasing aft wall angle reduces the cavity drag due to the stable flow field which is vital for flame holding in supersonic combustor. Uniform mixing and gradual decrease in stagnation pressure loss can be achieved by decreasing the cavity aft wall angle.

  4. The strength of the reactor cavity of VVER-1000 NPP against steam explosion

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1995-01-01

    The reactor cavity of VVER-1000 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. The static analysis of the structure used the ABAQUS/STANDARD and ANSYS codes. The material properties in both runs were specified to be elasto-plastic, and the cracking of concrete was taken into account. (author). 2 refs., 5 figs

  5. 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

  6. 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

  7. Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

    International Nuclear Information System (INIS)

    Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

    2015-01-01

    Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

  8. 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

  9. 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)

  10. 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

  11. 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

  12. Radiation Damage In Reactor Cavity Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G [ORNL; Le Pape, Yann [ORNL; Naus, Dan J [ORNL; Remec, Igor [ORNL; Busby, Jeremy T [ORNL; Rosseel, Thomas M [ORNL; Wall, Dr. James Joseph [Electric Power Research Institute (EPRI)

    2015-01-01

    License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has established a renewed focus on long-term aging of nuclear generating stations materials, and recently, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis (EMDA), jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete. Much of the historical mechanical performance data of irradiated concrete does not accurately reflect typical radiation conditions in NPPs or conditions out to 60 or 80 years of radiation exposure. To address these potential gaps in the knowledge base, The Electric Power Research Institute and Oak Ridge National Laboratory are working to disposition radiation damage as a degradation mechanism. This paper outlines the research program within this pathway including: (i) defining the upper bound of the neutron and gamma dose levels expected in the biological shield concrete for extended operation (80 years of operation and beyond), (ii) determining the effects of neutron and gamma irradiation as well as extended time at temperature on concrete, (iii) evaluating opportunities to irradiate prototypical concrete under accelerated neutron and gamma dose levels to establish a conservative bound and share data obtained from different flux, temperature, and fluence levels, (iv) evaluating opportunities to harvest and test irradiated concrete from international NPPs, (v) developing cooperative test programs to improve confidence in the results from the various concretes and research reactors, (vi) furthering the understanding of the effects of radiation on concrete (see companion paper) and (vii) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge.

  13. 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

  14. 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.

  15. The effect of water vapor in the reactor cavity in a MHTGR [Modular High Temperature Gas Cooled Reactor] on the radiation heat transfer

    International Nuclear Information System (INIS)

    Cappiello, M.W.

    1991-01-01

    Analyses have been completed to determine the effect of the presence of water vapor in the reactor cavity in a modular high temperature gas cooled reactor on the predicted radiation heat transfer from the vessel wall to the reactor cavity cooling system. The analysis involves the radiation heat transfer between two parallel plates with an absorbing and emitting medium present. Because the absorption in the water vapor is spectrally dependent, the solution is difficult even for simple geometries. A computer code was written to solve the problem using the Monte Carlo method. The code was validated against closed form solutions, and shows excellent agreement. In the analysis of the reactor problem, the results show that the reduction in heat transfer, and the consequent increase in the vessel wall temperature, can be significant. This effect can be cast in terms of a reduction in the wall surface emissivities from 0.8 to 0.59. Because of the insulating effect of the water vapor, increasing the gap distance between the vessel wall and the cooling system will cause the vessel wall temperature to increase further. Care should be taken in the design of the facility to minimize the gap distance and keep temperature increase within allowable limits. 3 refs., 6 figs., 4 tabs

  16. 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

  17. 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

  18. Fabrication technology for a series of cylindrical thin-wall cavity targets

    CERN Document Server

    Zheng Yong; Sun Zu Oke; Wang Ming Da; Zhou La; Zhou Zhi Yun

    2002-01-01

    Cylindrical thin-wall cavity targets have been fabricated to study the behavior of superthermal electrons and their effects on inertial confinement fusion (ICF). Self-supporting cavity targets having adjustable, uniform wall thickness, and low surface roughness were required. This required production of high-quality mandrels, coating them by sputtering or electroplating, developing techniques for measurement of wall thickness and other cavity parameters, improving the uniformity of rotation of the mandrels, and preventing damage to the targets during removal from the mandrels. Details of the fabrication process are presented. Experimental results from the use of these targets are presented. These results, in good agreement with simulations, indicate that the use of thin-wall cavity targets is an effective method for studying superthermal electrons in ICF.

  19. 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

  20. 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. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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.)

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  6. 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

  7. 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

  8. 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.)

  9. 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)

  10. 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

  11. Double-walled tank type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Nishiguchi, Yohei.

    1993-01-01

    A secondary vessel containing a steam generator is disposed on a base slab, and a roof slab is disposed to the upper end opening of the base slab. A manometer sealing is formed between the upper end opening of the secondary vessel and the roof slab. A primary vessel is disposed in the second vessel for containing a reactor core therein. A communication pipeline system (equalizer) is disposed for communicating the cover gas space of the secondary vessel with the cover gas space of the primary vessel by way of the roof slab. The communication pipeline system comprises a breakable plate, a check valve which opens from the secondary system to the primary system, a closing valve and pipelines connecting them. Upon occurrence of a sodium-water reaction accident caused by rupture of heat transfer pipes of a steam generator in the secondary vessel, the breakable plate is broken to equalize the gas pressure by way of the communication pipelines. This can avoid external pressure buckling of the primary vessel. (I.N.)

  12. 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

  13. 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

  14. 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

  15. 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.

  16. Nuclear reactor assembly

    International Nuclear Information System (INIS)

    Dorner, H.; Scholz, M.; Jungmann, A.

    1975-01-01

    A nuclear reactor assembly includes a reactor pressure tank having a substantially cylindrical side wall surrounded by the wall of a cylindrical cavity formed by a biological shield. A rotative cylindrical wall is interposed between the walls and has means for rotating it from outside of the shield, and a probe is carried by the rotative wall for monitoring the pressure tank's wall. The probe is vertically movable relative to the rotative cylindrical wall, so that by the probe's vertical movement and rotation of the rotative cylinder, the reactor's wall can be very extensively monitored. If the reactor pressure tank's wall fails, it is contained by the rotative wall which is backed-up by the shield cavity wall. (Official Gazette)

  17. 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

  18. DEMOS PLUS. Robot for decontaminating soils and cavity walls of the reactor and fuel pools NPP primarily during periods of recharging fuel; DEMOS PLUS. Robot para la descontaminacion de suelos y paredes de la cavidad de reactor y piscinas de combustible de CC.NN. principalmente durante los periodos de recarga de combustible

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

    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)

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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)

  17. Plasma core reactor applications

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  18. 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.

  19. 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

  20. 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

  1. 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)

  2. 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.

  3. 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

  4. 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.)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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.)

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. 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)…

  20. 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

  1. 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.

  2. 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.

  3. Poloidal variations in toroidal fusion reactor wall power loadings

    International Nuclear Information System (INIS)

    Carroll, M.C.; Miley, G.H.

    1985-01-01

    A geometric formulation is developed by the authors for determining poloidal variations in bremmstrahlung, cyclotron radiation, and neutron wall power loadings in toroidal fusion devices. Assuming toroidal symmetry and utilizing a numerical model which partitions the plasma into small cells, it was generally found that power loadings are highest on the outer surface of the torus, although variations are not as large as some have predicted. Results are presented for various plasma power generation configurations, plasma volume fractions, and toroidal aspect ratios, and include plasma and wall blockage effects

  4. Leakage tests of wall segments of reactor containments

    International Nuclear Information System (INIS)

    Rizkalla, S.H.; Simmonds, S.H.; MacGregor, J.G.

    1979-10-01

    Two prestressed concrete wall segments simulating portions of containment walls were loaded by axial tensile forces to cause cracking of the concrete. At each load increment air pressure was applied in steps up to 21 psi to one side of the segment and the rate of leakage of air through the cracked concrete section was measured. A theoretical equation for the flow of air through concrete cracks is developed and the results from one leakage test are used to determine the dimensionless constant required for this equation. (author)

  5. 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.)

  6. 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

  7. 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)

  8. First-wall/blanket materials selection for STARFIRE tokamak reactor

    International Nuclear Information System (INIS)

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed

  9. 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.)

  10. 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

  11. Can wall and limiter erosion be eliminated in fusion reactors

    International Nuclear Information System (INIS)

    Norem, J.H.

    1981-10-01

    A pump limiter system is described which is compatible with in-situ recoating of the limiter surface. The recoating could be done during normal tokamak operation. We have shown how this system is compatible with most of the constraints of fusion reactor operation and might provide a significant advantage over magnetic diverter and some other pump limiter geometries

  12. 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

  13. 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.

  14. 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

  15. 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

  16. Absolute quantitation of left ventricular wall and cavity parameters using ECG-gated PET

    DEFF Research Database (Denmark)

    Freiberg, Jacob; Hove, Jens D; Kofoed, Klaus F

    2004-01-01

    in a heart phantom and in healthy subjects. Twelve healthy men aged 64 +/- 8 years were studied by use of cine magnetic resonance imaging (MRI) and ECG-gated FDG-PET during euglycemic glucose-insulin clamp. At increasing image noise levels, the estimated cavity volume of the heart phantom was within 2 m...

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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)

  2. Analysis of Cavity Pressure and Warpage of Polyoxymethylene Thin Walled Injection Molded Parts: Experiments and Simulations

    DEFF Research Database (Denmark)

    Guerrier, Patrick; Tosello, Guido; Hattel, Jesper Henri

    2014-01-01

    Process analysis and simulations on molding experiments of 3D thin shell parts have been conducted. Moldings were carried out with polyoxymethylene (POM). The moldings were performed with cavity pressure sensors in order to compare experimental process results with simulations. The warpage...... was characterized by measuring distances using a tactile coordinate measuring machine (CMM). Molding simulations have been executed taking into account actual processing conditions. Various aspects have been considered in the simulation: machine barrel geometry, injection speed profiles, cavity injection pressure......, melt and mold temperatures, material rheological and pvT characterization. Factors investigated for comparisons were: injection pressure profile, short shots length, flow pattern, and warpage. A reliable molding experimental database was obtained, accurate simulations were conducted and a number...

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

    Danilov, I.; Skladnov, K.; Kolganov, V.

    1994-01-01

    Previous designing of the first wall and limiter has allowed to determine their possible embodiment depending on the parameters and operation conditions of the blanket. As a rule limiter is a separate structure located on the plasma facing surface of the blanket assembly. Possible versions of the limiter/FW which may be considered: (1) limiters with mechanical attachment of the protective part; (2) limiters with the attachment with brazing; (3) limiters with common/separate cooling system; (4) limiter as a substitute of the FW. Generally the FW/limiter structure includes protective shield and its cooling system which consist of protective coating, heat accumulator, conductive layer and attachment locks

  9. 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

  10. 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.)

  11. 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

  12. 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)

  13. Transmutation and activation of fusion reactor wall and structural materials

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1979-01-01

    This report details the extent of the nuclear data needed for inclusion in a data library to be used for general assessments of fusion reactor structure activation and transmutation, describes the sources of data available, reviews the literature and explores the reliability of current calculations by providing an independent assessment of the activity inventory to be expected from five structural materials in a simple blanket design for comparison with the results of other workers. An indication of the nuclear reactions which make important contributions to the activity, transmutation and gas production rates for these structural materials is also presented. (author)

  14. 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.

  15. 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

  16. 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

  17. The moisture conditions of nuclear reactor concrete containment walls - an example for a BWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, L.O.; Johansson, P. [Lund Institute of Technology, Laboratory of Building Materials, PO Box 118, 221 00 Lund (Sweden)

    2006-07-01

    A method is presented on how to quantify the moisture conditions of nuclear concrete containment walls. The method is based on first quantifying the boundary conditions at the outer and inner surfaces and then describing the moisture fixation and moisture transport within the concrete wall. The temperature and humidity conditions of the outdoor air and of the air close to the wall surfaces are monitored for a period of time and the vapour contents in the different points are compared. From the differences between the vapour contents the sources of moisture are identified and quantified. The previous and future climatic conditions are then predicted. An example is given for the conditions in the containment walls at Barsebaeck nuclear power plant, where moisture measurements have been performed in situ and on samples taken from the walls. (authors)

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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.)

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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/

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. 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.)

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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)

  9. 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.)

  10. 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.

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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.)

  1. 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.

  2. 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

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Huh, Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)] (and others)

    1999-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the Nuclear Power Plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the second phase of this project, current status of research about corium-concrete interaction and corium coolability which can occur on the reactor cavity has been surveyed, and the researches about lower head failure mechanism have also been surveyed. And, severe accident analysis for Ulchin 3 and 4 has been conducted, and collapse load of lower head has been analyzed through structural analysis considering various heat transfer conditions. The results of accident analysis can be used as a basic input for structural analysis which will be conducted in 3rd phase of this study.

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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.)

  10. Destruction of an industrial wastewater by supercritical water oxidation in a transpiring wall reactor

    International Nuclear Information System (INIS)

    Bermejo, M.D.; Cocero, M.J.

    2006-01-01

    The supercritical water oxidation (SCWO) is a technology that takes advantage of the special properties of water in the surroundings of critical point of water to completely oxidize wastes in residence times lower than 1 min. The problems caused by the harsh operational conditions of the SCWO process are being solved by new reactor designs, such as the transpiring wall reactor (TWR). In this work, the operational parameters of a TWR have been studied for the treatment of an industrial wastewater. As a result, the process has been optimized for a feed flow of 16 kg/h with feed inlet temperatures higher than 300 deg. C and transpiring flow relation (R) between 0.2 and 0.6 working with an 8% (w/w) isopropanol (IPA) as a fuel. The experimental data and a mathematical model have been applied for the destruction of an industrial waste containing acetic acid and crotonaldehyde as main compounds. As the model predicted, removal efficiencies higher than 99.9% were obtained, resulting in effluents with 2 ppm total organic carbon (TOC) at feed flow of 16 kg/h, 320 deg. C of feed temperature and R = 0.32. An effluent TOC of 35 ppm under conditions feed flow of 18 kg/h, feed inlet temperatures of 290 deg. C, reaction temperatures of 570 deg. C and R = 0.6

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.

  16. Labyrinths, columns and cavities: new internal features of pollen grain walls in the Acanthaceae detected by FIB-SEM.

    Science.gov (United States)

    House, Alisoun; Balkwill, Kevin

    2016-03-01

    External pollen grain morphology has been widely used in the taxonomy and systematics of flowering plants, especially the Acanthaceae which are noted for pollen diversity. However internal pollen wall features have received far less attention due to the difficulty of examining the wall structure. Advancing technology in the field of microscopy has made it possible, with the use of a focused ion beam-scanning electron microscope (FIB-SEM), to view the structure of pollen grain walls in far greater detail and in three dimensions. In this study the wall structures of 13 species from the Acanthaceae were investigated for features of potential systematic relevance. FIB-SEM was applied to obtain precise cross sections of pollen grains at selected positions for examining the wall ultrastructure. Exploratory studies of the exine have thus far identified five basic structural types. The investigations also show that similar external pollen wall features may have a distinctly different internal structure. FIB-SEM studies have revealed diverse internal pollen wall features which may now be investigated for their systematic and functional significance.

  17. 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.

  18. 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.

  19. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

    International Nuclear Information System (INIS)

    Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

    2005-01-01

    The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m 2 .s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m 2 .s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux ( 2 .s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water

  20. Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

    International Nuclear Information System (INIS)

    Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

    1987-01-01

    The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

  1. 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.

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

    Science.gov (United States)

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

    1988-01-01

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

  3. 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

  4. 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.)

  5. 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

  6. Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

    Energy Technology Data Exchange (ETDEWEB)

    Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States); Petrov, Victor [Univ. of Michigan, Ann Arbor, MI (United States); Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Tompkins, Casey [Univ. of Wisconsin, Madison, WI (United States); Nunez, Daniel [Univ. of Michigan, Ann Arbor, MI (United States)

    2018-02-13

    This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has been developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.

  7. 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

  8. 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.

  9. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  10. Effects of torus wall flexibility on forces in the Mark I Boiling Water Reactor Pressure Suppression System. Part I

    International Nuclear Information System (INIS)

    Martin, R.W.; McCauley, E.W.

    1977-09-01

    The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems

  11. Modified divergence theorem for analysis and optimization of wall reflecting cylindrical UV reactor

    Directory of Open Access Journals (Sweden)

    Milanović Đurđe R.

    2011-01-01

    Full Text Available In this paper, Modified Divergence Theorem (MDT, known in earlier literature as Gauss-Ostrogradsky theorem, was formulated and proposed as a general approach to electromagnetic (EM radiation, especially ultraviolet (UV radiation reactor modeling. Formulated mathematical model, based on MDT, for multilamp UV reactor was applied to all sources in a reactor in order to obtain intensity profiles at chosen surfaces inside reactor. Applied modification of MDT means that intensity at a real opaque or transparent surface or through a virtual surface, opened or closed, from different sides of the surface are added and not subtracted as in some other areas of physics. Derived model is applied to an example of the multiple UV sources reactor, where sources are arranged inside a cylindrical reactor at the coaxial virtual cylinder, having the radius smaller than the radius of the reactor. In this work, optimization of a reactor means maximum transfer of EM energy sources into the fluid for given fluid absorbance and fluid flow-dose product. Obtained results, for in advanced known water quality, gives unique solution for an optimized model of a multilamp reactor geometry. As everyone can easily verify, MDT is very good starting point for every reactor modeling and analysis.

  12. Reactor cavity seal ring

    International Nuclear Information System (INIS)

    Hankinson, M.F.

    1986-01-01

    A hydrostatic seal is described for sealing an annular gap between two flat substantially horizontal coplanar surfaces comprising, in combination: a generally flat annular plate of a width sufficient to span a gap between two surfaces: compressible annular sealing means disposed on the bottom surface of the flat annular plate for sealingly engaging the two flat surfaces in response to a downward force exerted on the plate; and fastening means, distributed along the center line of the plate, for releasably fastening the plate in a position to span the gap to be sealed and exert a downward force on the plate, each fastening means including a pair of elongated members of a size to fit into the gap to be sealed, means for mounting the members on the bottom surface of the plate so that at least a portion of each member is radially moveable in a direction toward a respective one of the vertical side surfaces defining the gap to be sealed to engage same and so that the plate is moveable relative to the members in a downward direction in response to hydrostatic pressure applied to the upper surface of the plate when the members are engaging the vertical side surfaces of an annular gap, and an actuating means, mounted on the plate for movement therewith in response to hydrostatic pressure, for radially moving the members, the actuating means extending through a bore in the plate to the upper surface of the plate

  13. 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.)

  14. 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

  15. 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)

  16. 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.

  17. 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

  18. 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

  19. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    International Nuclear Information System (INIS)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

  20. 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

  1. 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.

  2. 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

  3. 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

  4. Perfusion of surgical cavity wall enhancement in early post-treatment MR imaging may stratify the time-to-progression in glioblastoma.

    Directory of Open Access Journals (Sweden)

    Ji Eun Park

    Full Text Available To determine if perfusion in surgical cavity wall enhancement (SCWE obtained in early post-treatment MR imaging can stratify time-to-progression (TTP in glioblastoma.This study enrolled 60 glioblastoma patients with more than 5-mm-thick SCWEs as detected on contrast-enhanced MR imaging after concurrent chemoradiation therapy. Two independent readers categorized the shape and perfusion state of SCWEs as nodular or non-nodular and as having positive or negative perfusion compared with the contralateral grey matter on arterial spin labeling (ASL. The perfusion fraction on ASL within the contrast-enhancing lesion was calculated. The independent predictability of TTP was analyzed using the Kaplan-Meier method and Cox proportional hazards modelling.The perfusion fraction was higher in the non-progression group, significantly for reader 2 (P = 0.03 and borderline significantly for reader 1 (P = 0.08. A positive perfusion state and (P = 0.02 a higher perfusion fraction of the SCWE were found to become an independent predictor of longer TTP (P = 0.001 for reader 1 and P < 0.001 for reader 2. The contrast enhancement pattern did not become a TTP predictor.Assessment of perfusion in early post-treatment MR imaging can stratify TTP in patients with glioblastoma for adjuvant temozolomide therapy. Positive perfusion in SCWEs can become a predictor of a longer TTP.

  5. Hot isostatically pressed (HIPed) thick-walled component for a pressurised water reactor (PWR) application

    International Nuclear Information System (INIS)

    Hookham, I.; Burdett, B.; Bridger, K.; Sulley, J.L.

    2009-01-01

    This paper presents the work conducted to justify and provide a quality assured HIPed thick-walled component for a PWR application; the component being designed and manufactured by Rolls-Royce. Rolls-Royce has previously published (ICAPP 08) its overall, staged approach to the introduction of powder HIPed components; starting with thin-walled, leak limited pressure boundaries, and culminating in the use of the powder HIPed process for thick walled components. This paper presents details specific to a thick walled pressure vessel component. Results are presented of non-destructive and destructive examinations of one of a batch of components. Mechanical testing and metallurgical examination results of sample material taken from different sections of the component are presented. A full range of test results is provided covering, as examples: tensile, Charpy impact and sensitization susceptibility. Differences in weldability between the HIPed and the previous forged form are also documented. (author)

  6. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    International Nuclear Information System (INIS)

    Corradin, Michael; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-01-01

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  7. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  8. 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.

  9. Synthesis of Multi-Walled Carbon Nanotubes from Plastic Waste Using a Stainless-Steel CVD Reactor as Catalyst

    Directory of Open Access Journals (Sweden)

    Pranav K. Tripathi

    2017-09-01

    Full Text Available The disposal of non-biodegradable plastic waste without further upgrading/downgrading is not environmentally acceptable and many methods to overcome the problem have been proposed. Herein we indicate a simple method to make high-value nanomaterials from plastic waste as a partial solution to the environmental problem. Laboratory-based waste centrifuge tubes made of polypropylene were chosen as a carbon source to show the process principle. In the process, multi-walled carbon nanotubes (MWCNTs were synthesized from plastic waste in a two-stage stainless steel 316 (SS 316 metal tube that acted as both reactor vessel and catalyst. The steel reactor contains Fe (and Ni, and various alloys, which act as the catalyst for the carbon conversion process. The reaction and products were studied using electron probe microanalysis, thermogravimetric analysis, Raman spectroscopy and transmission electron microscopy and scanning electron microscopy. Optimization studies to determine the effect of different parameters on the process showed that the highest yield and most graphitized MWCNTs were formed at 900 °C under the reaction conditions used (yield 42%; Raman ID/IG ratio = 0.48. The high quality and high yield of the MWCNTs that were produced in a flow reactor from plastic waste using a two stage SS 316 chemical vapor deposition (CVD furnace did not require the use of an added catalyst.

  10. Synthesis of Multi-Walled Carbon Nanotubes from Plastic Waste Using a Stainless-Steel CVD Reactor as Catalyst.

    Science.gov (United States)

    Tripathi, Pranav K; Durbach, Shane; Coville, Neil J

    2017-09-22

    The disposal of non-biodegradable plastic waste without further upgrading/downgrading is not environmentally acceptable and many methods to overcome the problem have been proposed. Herein we indicate a simple method to make high-value nanomaterials from plastic waste as a partial solution to the environmental problem. Laboratory-based waste centrifuge tubes made of polypropylene were chosen as a carbon source to show the process principle. In the process, multi-walled carbon nanotubes (MWCNTs) were synthesized from plastic waste in a two-stage stainless steel 316 (SS 316) metal tube that acted as both reactor vessel and catalyst. The steel reactor contains Fe (and Ni, and various alloys), which act as the catalyst for the carbon conversion process. The reaction and products were studied using electron probe microanalysis, thermogravimetric analysis, Raman spectroscopy and transmission electron microscopy and scanning electron microscopy. Optimization studies to determine the effect of different parameters on the process showed that the highest yield and most graphitized MWCNTs were formed at 900 °C under the reaction conditions used (yield 42%; Raman I D / I G ratio = 0.48). The high quality and high yield of the MWCNTs that were produced in a flow reactor from plastic waste using a two stage SS 316 chemical vapor deposition (CVD) furnace did not require the use of an added catalyst.

  11. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Bloom, E.E.

    1979-01-01

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,..cap alpha..) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures.

  12. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.

    1979-01-01

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,α) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures

  13. Probabilistic methods for evaluation of erosion-corrosion wall thinning in french pressurized water reactors

    International Nuclear Information System (INIS)

    Ardillon, E.; Bouchacourt, M.

    1994-04-01

    This paper describes the application of the probabilistic approach to a selected study section having known characteristics. The method is based on the physico-chemical model of erosion-corrosion, the variables of which are probabilized. The three main aspects of the model, namely the thermohydraulic flow conditions, the chemistry of the fluid, and the geometry of the installation, are described. The study ultimately makes it possible determine: - the evolution of wall thinning distribution, using the power station's measurements; - the main parameters of influence on the kinetics of wall thinning; - the evolution of the fracture probabilistic of the pipe in question. (authors). 10 figs., 7 refs

  14. Evaluation of thermal ratcheting of reactor vessel wall near the sodium surface

    International Nuclear Information System (INIS)

    Take, Kohji; Fujioka, Terutaka; Yano, Kazutaka

    1989-01-01

    Plastic ratcheting of reactor vessels may occur by an axially moving thermal gradient without primary stress. So there is a need to establish a proper prediction method for the plastic ratcheting. In this study, inelastic FEM analyses of reactor vessel model by using an advanced constitutive equation were carried out in order to comprehend plastic ratcheting behaviour of cylinder which subject to an axially moving thermal gradient. As a result of analyses, a basic mechanism of this ratcheting was found. And it also indicated that cyclic hardening behaviour will became important for development of evaluation method. (author)

  15. Effect of plasma physics on choices of first wall materials and structures for a thermonuclear reactor

    International Nuclear Information System (INIS)

    Meade, D.M.

    1975-01-01

    Impurity ions adversely affect the behavior of present-day tokamaks, and control of impurities is expected to be a key element in determining the feasibility of thermonuclear fusion reactors. The plasma-surface interactions for tokamaks and several techniques for controlling impurities are described. The plasma-surface problem of next generation devices PLT, PDX, DIII and TFTR is expected to be similar to those encountered in a reactor. For these devices calculations indicate that most of the particle energy efflux will be in the 1 keV region. Ironically this energy region has not yet been investigated thoroughly by the surface physicists

  16. Review of the works on plasma-wall interactions in fusion reactors, 1

    International Nuclear Information System (INIS)

    Sone, Kazuho

    1975-09-01

    A review is made of the works on sputtering as one of the plasma-wall interactions in thermonuclear fusion devices. The present status and future problems are described mainly in experiments of low-energy light ions such as H + , H 2 + , D + , D 2 + and He + , heavy ions including self-ions, and fast neutrons for polycrystalline metal targets. (auth.)

  17. Evaluation of copper alloys for fusion reactor divertor and first wall components

    DEFF Research Database (Denmark)

    Fabritsiev, S.A.; Zinkle, S.J.; Singh, B.N.

    1996-01-01

    This paper presents a critical analysis of the main factors of radiation damage limiting the possibility to use copper alloys in the ITER divertor and first wall structure. In copper alloys the most significant types of radiation damage in the proposed temperature-dose operation range are swellin...

  18. 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

  19. 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.

  20. 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

  1. 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.

  2. ICRF boronization. A new technique towards high efficiency wall coating for superconducting tokamak reactors

    International Nuclear Information System (INIS)

    Li Jiangang; Zhao Yan Ping; Gu Xue Mao

    1999-01-01

    A new technique for wall conditioning that will be especially useful for future larger superconducting tokamaks, such as ITER, has been successfully developed and encouraging results have been obtained. Solid carborane powder, which is non-toxic and non-explosive, was used. Pulsed RF plasma was produced by a non-Faraday shielding RF antenna with RF power of 10 kW. The ion temperature was about 2 keV with a toroidal magnetic field of 1.8 T and a pressure of 3x10 -1 Pa. Energetic ions broke up the carborane molecules, and the resulting boron ions struck and were deposited on the first wall. In comparison with glow discharge cleaning boronization, the B/C coating film shows higher adhesion, more uniformity and longer lifetime during plasma discharges. The plasma performance was improved after ICRF boronization. (author). Letter-to-the-editor

  3. A preliminary model for estimating the first wall lifetime of a fusion reactor

    International Nuclear Information System (INIS)

    Daenner, W.

    1975-02-01

    The estimation of the first wall lifetime is a necessary basis for predicting the availability of a fusion power plant. In order to do this, an analytical model was prepared and programmed for the computer which calculates the temperature and stress load of the first wall from the principal design parameters and quotes them against the relevant material properties. Neither the analytical model nor the information about the material performance is yet complete so that the answers obtained from the program are very preliminary. This situation is underlined by the results of sample calculations performed for the CTRD blanket module cell. The results obtained for vanadium and vanadium alloys show a strong dependence of the lifetime on the irradiation creep and the ductility of these materials. Completion of this model is envisaged as soon as the missing information becomes available. (orig.) [de

  4. 'Thin walled' concept and a new top lid applied to the Scandinavian PCRV for a boiling water reactor

    International Nuclear Information System (INIS)

    Scotto, F.L.

    1975-01-01

    This research is carried out in the frame of an agreement between AB ATOMENERGI of Sweden and ENEL (Ente Nazionale per l'Energia Elettrica) of Italy, for an exchange of information in the field of PCPV for BWR, and takes as a reference the Scandinavian solution as far as the thermal insulation system and the geometry are concerned, proposing new solutions for the prestressed concrete structure (namely the Author's concept of thin walls and a new concept of top lid). The proposed top lid sealing system solution is in line with the one adopted for the conventional steel pressure vessel enclosures; furthermore the prestressed concrete lid is restricted to the prestressed concrete structure to form a continuous contrete structure, in line with th PCPV conventional solutions for gas reactors. The paper describes in detail the selected design philosophy that is slightly different from the one defined by the Scandinavian project. In fact, as far as the design limits are concerned, it refers mainly to steel pressure vessel philosophy and, as to the concrete behaviour, to the design philosophy proposed by the author for the PCPV 'thin walled' structures for gas-cooled power reactors. Rheological, mathematical and physical models had been suitably devised in order to check the reliability of the proposed assumption. This paper therefore, will also give a brief description of said tools and the main results acquired at the time of the conference, and technical and economical considerations made to support the interest of the research, showing the relevant cut down of the costs. The comparative reference steel pressure vessel belongs to Mark III ENEL VI and VIII BW plant to which design and construction the author gives his contribution

  5. Some initial considerations on the suitability of Ferritic/ martensitic stainless steels as first wall and blanket materials in fusion reactors

    International Nuclear Information System (INIS)

    Butterworth, G.J.

    1982-01-01

    The constitution of stainless iron alloys and the characteristic properties of alloys in the main ferritic, martensitic and austenitic groups are discussed. A comparison of published data on the mechanical, thermal and irradiation properties of typical austenitic and martensitic/ferritic steels shows that alloys in the latter groups have certain advantages for fusion applications. The ferromagnetism exhibited by martensitic and ferritic alloys has, however, been identified as a potentially serious obstacle to their utilisation in magnetic confinement devices. The paper describes measurements performed in other laboratories on the magnetic properties of two representative martensitic alloys 12Cr-1Mo and 9Cr-2Mo. These observations show that a modest bias magnetic field of magnitude 1 - 2 tesla induces a state of magnetic saturation in these materials. They would thus behave as essentially paramagnetic materials having a relative permeability close to unity when saturated by the toroidal field of a tokamak reactor. The results of computations by the General Atomic research group to assess the implications of such magnetic behaviour on reactor design and operation are presented. The results so far indicate that the ferromagnetism of martensitic/ferritic steels would not represent a major obstacle to their utilisation as first wall or blanket materials. (author)

  6. Low cycle fatigue lifetime of HIP bonded Bi-metallic first wall structures of fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hashimoto, Toshiyuki; Kitamura, Kazunori

    1998-10-01

    A HIP bonded bi-metallic panel composed of a dispersion strengthened copper (DSCu) layer and type 316L stainless steel (SS316L) cooling pipes is the reference design of the ITER first wall. To examine the fatigue lifetime of the first wall panel under cyclic mechanical loads, low cycle fatigue tests of HIP bonded bi-metallic specimens made of SS316L and DSCu were conducted with the stress ratio of -1.0 and five nominal strain range conditions ranging from 0.2 to 1.0%. Elasto-plastic analysis has also been conducted to evaluate local strain ranges under the nominal strains applied. Initial cracks were observed at the inner surface of the SS316L cooling pipes for all of the specimens tested, which was confirmed by the elasto-plastic analysis that the maximum strains of the test specimens were developed at the same locations. It was found that the HIP bonded bi-metallic test specimens had a fatigue lifetime longer than that of the SS316L raw material obtained by round bar specimens. Similarly, the fatigue lifetime of the DSCu/SS316L HIP interface was also longer than the round bar test results for the HIP joints. From these results, it has been confirmed that the bi-metallic first wall panel with built-in cooling pipes made by HIP bonding has a sufficient fatigue lifetime in comparison with the raw fatigue data of the materials, which also suggests that the fatigue lifetime evaluation has an adequate margin against fracture if it follows the design fatigue curve based on the material fatigue data. (author)

  7. Air filtration plants of wall-type for separation of fission iodine in nuclear reactors

    International Nuclear Information System (INIS)

    Stiehl, H.H.; Neumann, M.; Sinhuber, D.

    1976-01-01

    The increasing density of nuclear power stations and increased safety requirements will lead in future to higher flow rates and longer residence times in the adsorption filter layer of the iodine sorption filter plants of nuclear power stations. The safety requirements in the Federal Republic of Germany have been complied with so far in the conventional way by means of duct-type filter constructions. For the higher flow rates and longer residence times necessary in future, we propose a filter construction of wall-type, which complies with the safety regulations of the Federal Republic of Germany. The economic and technical advantages are discussed

  8. 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.

  9. Scientific report. Plasma-wall interaction studies related to fusion reactor materials

    International Nuclear Information System (INIS)

    Temmerman, G. De

    2006-01-01

    This scientific report summarises research done on erosion and deposition mechanisms affecting the optical reflectivity of potential materials for use in the mirrors used in fusion reactors. Work done in Juelich, Germany, at the Federal Institute of Technology in Lausanne, Switzerland, the JET laboratory in England and in Basle is discussed. Various tests made with the mirrors are described. Results obtained are presented in graphical and tabular form and commented on. The influence of various material choices on erosion and deposition mechanisms is discussed

  10. Thermoelectric conversion at the divertor plates and the first wall of a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, S. [National Inst. for Fusion Science, Nagoya (Japan); Sagara, A. [National Inst. for Fusion Science, Nagoya (Japan); Komori, A. [National Inst. for Fusion Science, Nagoya (Japan); Tazima, T. [National Inst. for Fusion Science, Nagoya (Japan); Motojima, O. [National Inst. for Fusion Science, Nagoya (Japan); Iiyoshi, A. [National Inst. for Fusion Science, Nagoya (Japan); Matsubara, K. [National Inst. for Fusion Science, Nagoya (Japan)]|[Yamaguchi Univ. (Japan); Onozuka, M. [National Inst. for Fusion Science, Nagoya (Japan)]|[Mitsubishi Heavy Industries Ltd. (Japan); Koganezawa, K. [National Inst. for Fusion Science, Nagoya (Japan)]|[Mitsubishi Heavy Industries Ltd. (Japan); Matsuda, T. [National Inst. for Fusion Science, Nagoya (Japan)]|[Toyo Tanso Co. Ltd. (Japan)

    1995-12-31

    We investigated thermoelectric conversion on the first wall and the divertor plates. Carbon, B{sub 4}C, and other carbon-based materials were tested as components of a thermoelectric element. The heat flux from the plasma was assumed to be 400 kW/m{sup 2}, and the cooling side temperature the fixed design parameter of either 350 K or 650 K. While differential radiation cooling was not considered in this study, a computer programme was used to estimate the distribution of temperature and thermal stress over the thermoelectric element. The three-legged element was conceived to be 20 cm long and 12 cm wide. The temperature in its arches reached almost 2500 K, and the maximal thermal stress was 80 MPa - still within the acceptable range for the ITER design parameter. The high thermoelectric power of B{sub 4}C accounts for the thermal efficiency of 2.8% (for 650 K) or 3.3% (for 350 K). If we find an N-type semi-conductor material with the same high absolute value as B{sub 4}C to replace carbon, the efficiency will improve to 9.4% (for 650 K) or 11% (for 350 K). Since plasma is a current-conducting medium, we discuss aspects of a plasma-connected thermoelectric element. Its efficiency would depend on the connection length of magnetic field and plasma parameters near the wall. (orig.).

  11. Applied model of through-wall crack of coolant vessels of WWER-type reactors

    International Nuclear Information System (INIS)

    Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

    2010-01-01

    We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

  12. 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.)

  13. 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.

  14. Effect of off-normal events on the reactor first wall

    International Nuclear Information System (INIS)

    Igitkhanov, Yu; Bazylev, B

    2011-01-01

    In this paper, we analyse the energy deposition and erosion of the W/EUROFER blanket module for the first wall (FW) of DEMO due to the runaway electrons (RE) and vertical displacements events (VDEs). The DEMO data for transients were extrapolated from ITER data by using the scaling arguments. The simulations were performed at an RE deposition energy in the range 30-100 MJ m - 2 over 0.05-0.3 s. In the case of a 'hot' VDE, all stored plasma energy is deposited on the FW area for ∼1 s. For a VDE following the thermal quench phase the remaining magnetic energy is deposited on the FW for ∼0.3 s. It is shown that the minimum W thickness needed for preventing failure of the W/EUROFER bond (assumed to be the EUROFER creep point) is large enough, causing armour melting. Both RE and VDE in DEMO will pose a major life-time issue depending on their frequency.

  15. Effect of off-normal events on the reactor first wall

    Science.gov (United States)

    Igitkhanov, Yu; Bazylev, B.

    2011-12-01

    In this paper, we analyse the energy deposition and erosion of the W/EUROFER blanket module for the first wall (FW) of DEMO due to the runaway electrons (RE) and vertical displacements events (VDEs). The DEMO data for transients were extrapolated from ITER data by using the scaling arguments. The simulations were performed at an RE deposition energy in the range 30-100 MJ m-2 over 0.05-0.3 s. In the case of a 'hot' VDE, all stored plasma energy is deposited on the FW area for ~1 s. For a VDE following the thermal quench phase the remaining magnetic energy is deposited on the FW for ~0.3 s. It is shown that the minimum W thickness needed for preventing failure of the W/EUROFER bond (assumed to be the EUROFER creep point) is large enough, causing armour melting. Both RE and VDE in DEMO will pose a major life-time issue depending on their frequency.

  16. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  17. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    International Nuclear Information System (INIS)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage

  18. 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

  19. Chooz-B1, the new Electricite de France PWR: calculation scheme of neutron leakages from the reactor cavity

    International Nuclear Information System (INIS)

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

    1987-04-01

    A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable

  20. 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

  1. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    International Nuclear Information System (INIS)

    Tricot, N.; Jendrich, U.

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  2. Welding and cutting characteristics of blanket/first wall module to back plate for fusion experimental reactor

    International Nuclear Information System (INIS)

    Sato, Shinichi; Osaki, Toshio; Koga, Shinji

    1996-01-01

    The first wall and the blanket of the International Thermonuclear Experimental Reactor (ITER) are used under severe conditions such as the neutron irradiation by plasma, surface thermal load, the electromagnetic force at the time of plasma disruption and others. Consequently, from the viewpoint of the necessity for disassembling and maintenance, those are divided into modules in toroidal and poloidal directions. In this study, as to the welding of the back plate and the legs supporting blanket modules, which are installed in a vacuum vessel, the characteristic test paying attention to the deformation at the time of welding was carried out, and the optimal welding conditions and the characteristics of welding deformation and others were clarified. Moreover, when water jet method was used for cutting the welded parts of the supporting legs, the properties of the cut parts, the time for cutting and others were examined. The performance required for the welded parts of blanket modules with back plate is shown. The basic test of welding conditions using plate models, partial model test and whole model test are reported. The test of water jet cutting for the maintenance of shielding blanket modules is described. (K.I.)

  3. Residual stress in the first wall coating materials of TiC and TiN for fusion reactor

    International Nuclear Information System (INIS)

    Qiu Shaoyu

    1997-01-01

    Residual stresses measurement in the first wall coating of a fusion reactor of TiC and TiN films by X-ray diffraction 'sin 2 ψ methods' were described. The authors have studied on the effect of conditions of specimen preparation (such as coating method, substrate materials, film thickness and deposition temperature) on the residual stress of TiC and TiN films coated onto Mo, 316LSS and Pocographite by chemical vapor deposition (CVD) and physical vapor deposition (PVD) method. All films prepared in this study were found to have a compressive stresses and the CVD method gave lower residual stress than PVD method. TiC film coated on Mo substrate at 1100 degree C by CVD method showed that residual stress as the film thickness was raised from 14 μm to 60 μm, on the other hand, residual stress by PVD method exhibited a high compressive stresses, this kind of stress was principally the intrinsic stress, and a marked decrease in the residual with raising the deposition temperature (200 degree C∼650 degree C) was demonstrated. Origins of the residual stress were discussed by correlation with differences between thermal expansion coefficients, and also with fabrication methods

  4. Chemical vapour deposition of silicon under reduced pressure in a hot-wall reactor: Equilibrium and kinetics

    International Nuclear Information System (INIS)

    Langlais, F.; Hottier, F.; Cadoret, R.

    1982-01-01

    Silicon chemical vapour deposition (SiH 2 Cl 2 /H 2 system), under reduced pressure conditions, in a hot-wall reactor, is presented. The vapour phase composition is assessed by evaluating two distinct equilibria. The homogeneous equilibrium , which assumes that the vapour phase is not in equilibrium with solid silicon, is thought to give an adequate description of the vapour phase in the case of low pressure, high gas velocities, good temperature homogeneity conditions. A comparison with heterogeneous equilibrium enables us to calculate the supersaturation so evidencing a highly irreversible growth system. The experimental determination of the growth rates reveals two distinct temperature ranges: below 1000 0 C, polycrystalline films are usually obtained with a thermally activated growth rate (+40 kcal mole -1 ) and a reaction order, with respect to the predominant species SiCl 2 , close to one; above 1000 0 C, the films are always monocrystalline and their growth rate exhibits a much lower or even negative activation energy, the reaction order in SiCl 2 remaining about one. (orig.)

  5. 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

  6. 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

  7. 3. IAEA research co-ordination meeting on atomic and plasma-wall interaction data for fusion reactor divertor modeling. Summary report

    International Nuclear Information System (INIS)

    Janev, R.K.

    1999-04-01

    A brief description of the proceedings and the conclusions of the 3rd Research Co-ordination Meeting on 'Atomic and Plasma-Wall Interaction Data for Fusion Reactor Divertor Modeling', held on March 8-9, 1999, at the IAEA Headquarters in Vienna, Austria, is provided. The reports on the activities within the individual projects pertinent to the IAEA Co-ordinated Research program with the same title are given as appendix to the present report. (author)

  8. 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

  9. Dental cavities

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/001055.htm Dental cavities To use the sharing features on this page, please enable JavaScript. Dental cavities are holes (or structural damage) in the ...

  10. 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.)

  11. Evaluation of endcap welds in thin walled fuel elements of pressurised heavy water reactor by ultrasonic testing

    International Nuclear Information System (INIS)

    Subramanian, C.V.; Thavasimuthu, M.; Kalyansundaram, P.; Bhattacharya, D.K.; Raj, Baldev

    1992-01-01

    In the pressurised heavy water reactor systems of India, the fuel is encapsulated in thin-walled tubes (0.342 mm) closed with endcaps by resistance welding. The integrity of these fuel elements should be such that no fission gas leakage takes place during reactor operation. The quality control of the endcap welds needed to satisfy this requirement includes helium leak test and destructive metallographic test (on sample basis). This paper discusses the feasibility study that has been carried out in the author's laboratory to develop an immersion ultrasonic test method for evaluating the integrity of the endcap weld region. Through holes of various sizes (0.15mm, 0.2mm, 0.4mm diameter and 0.185mm and 0.342mm deep) were machined by spark erosion machining at the weld joints to simulate defects of various sizes. Line focussed probe of 10 MHz frequency was used for the testing. It was possible to detect clearly all the machined holes. Based on the above standardised procedure, further testing was done on endcap welds which were rejected during fabrication on account of showing leak rate of 3 x 10 -6 std. c.c/sec. or more during helium leak test. Though it was possible to get echoes from the natural defects in the rejected tubes with echo amplitude of 70%, the signal was accompanied by the geometrical reflection (noise) giving an amplitude of 20% from the weld region, giving rise to the problem of resolving the defect indication from the geometric indications. Therefore, signal analysis approach was adopted. The signal obtained from the weld zone were subjected to various analysis procedures like a) autopower spectrum, b) total energy content and c) demodulated auto correlation function. It was possible by all the three methods to differentiate the defect signal from those due to weld geometry or due to noise. Subsequently, metallography was carried out to characterise the type of defects observed during the ultrasonic testing. (author). 4 figs

  12. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  13. Expanding the applicable duration for shrink fitting of the ultrathin-walled reactor coolant pump rotor-can

    International Nuclear Information System (INIS)

    Li, Ruiqin; Zhang, Chi; Zhang, Liwen; Cui, Yan; Shen, Wenfei

    2017-01-01

    Highlights: •A thermal-mechanical coupled finite element model was developed to simulate the whole process. •Heat capacity added layer was used to extend the limited time for the process. •Shrink-fitted experiments were performed to verify the simulation results. -- Abstract: The rotor-can of reactor coolant pump (RCP) is generally assembled on the rotor using shrink fitting technique. The rotor-can is characterized by large height and ultrathin-walled cylinder, thus, its rigidity is weak and heat capacity is quite limited. The shrink fitting process has to be completed within a short limited-time, which makes it difficult for rotor to insert in the rotor-can completely. In order to solve this problem, a new method was proposed to extend the limited time by using a heat capacity added layer (HCAL) during the shrink fitting process. A thermal-mechanical coupled finite element (FE) model was developed to simulate the whole process. The transient heat exchange with a narrow gap between rotor and rotor-can during the shrink fitting process was taken into consideration. The limited time was predicted by calculating and analyzing the evolutions of temperature field and radial displacement field of the rotor-can. The simulation results indicate that the limited time of the shrink fitting process can be significantly extended with the increase of HCAL in thickness. Then, shrink fitting experiments were performed to confirm the extending effect of the HCAL. The experimental results of limited time show good agreement with the predicted values. The current results will certainly help the designer to improve the shrink fitting technique.

  14. Fire barrier evaluation of the wall between spent nuclear fuel storage basins and reactor areas, 105KE and 105KW

    International Nuclear Information System (INIS)

    Johnson, B.H.

    1994-01-01

    This evaluation is intended to provide a comprehensive assessment of the risks from fire and fire related perils for the fire barriers between spent nuclear fuel storage basins and reactor areas, 105KE and 105KW. As a result of this fire barrier evaluation the present walls and the components thereof are not a true listed fire rated assembly. However, due to the construction of these barriers and the components thereof, these barriers will provide an equivalent level of protection provided the recommendations in Section 8.0 of this report are completed. These recommended upgrades are based upon sound engineering practice by a Registered Fire Protection Engineer. The construction of the barrier are substantial enough to provide the required 2-hr fire resistance rating. The primary concern is the numerous penetrations in the barrier. There are many penetrations that are adequate and no additional work is required. These penetrations are the ones that were poured-in-place at the time of construction. The penetrations that are of concern are some of the doors, the HVAC ducts, and the unsealed piping and conduit penetrations. There are several metal doors that should be replaced because the existing doors have either a non-approved window or louver that will not limit the spread of fire to one side of the barrier. All unsealed piping and conduit penetrations should be firestopped with an approved firestopping material. The existing non-active ducts that pass through the barrier should be disconnected at the barrier and the opening sealed with an approved firestopping method

  15. 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

  16. 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.

  17. 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

  18. An inter-comparison of HO2 measured by Fluorescence Assay by Gas Expansion and Cavity Ringdown Spectroscopy in the Highly Instrumented Reactor for Atmospheric Chemistry.

    Science.gov (United States)

    Brennan, A.; Onel, L. C.; Gianella, M.; Ronnie, G.; Aguila, A. L.; Hancock, G.; Whalley, L.; Seakins, P. W.; Ritchie, G.; Heard, D. E.

    2017-12-01

    HO2 is an important species in the atmosphere, as it is involved in the HOx radical reaction cycle that is critical to the oxidation of atmospheric pollutants and the ultimate cleaning of the troposphere. One of the most widely utilised methods to measure HO2 is Fluorescence Assay by Gas Expansion (FAGE), which indirectly measures HO2 by sampling into a low pressure cell and titrating HO2 with NO to produce OH that is then detected by Laser Induced Fluorescence. This is an indirect and non-absolute detection technique that requires careful calibration to convert the measured signal into [HO2], which involves the photolysis of H2O at 185 nm to produce OH and HO2, and is subject to 30 % errors at 2σ level. The work presented here shows the validation of the FAGE technique and its calibration procedure through inter-comparison experiments between the non-absolute FAGE technique and Cavity Ringdown Spectroscopy (CRDS), an absolute absorption based method. The CRDS system was used to excite the first O-H overtone of the HO2 absorption band at 1506.43 nm, and features a cavity length of 1.2 m and a total path of 60 km. The experiments were performed inside the 2.25 m3 stainless steel Highly Instrumented Reactor for Atmospheric Chemistry (HIRAC), using a synthetic air mixture at 150 and 1000 mbar of pressure and 298 K. HO2 was generated by photolysis of Cl2 at 365 nm in the presence of CH3OH and O2, and the [HO2] was monitored using both instruments. Additionally, monitoring the temporal decay of HO2 during its self-reaction provided an alternative calibration method for the FAGE instrument, and allowed the absorption cross section of HO2 at 1506.43 nm, σHO2, to be measured. FAGE calibration factors determined through the second order decays of HO2 at 1000 mbar agreed within 8 % of the H2O photolysis method, and determinations of σHO2 at 150 and 1000 mbar agree with previously reported data within 20 % and 12 % respectively. [HO2] correlation plots between the two

  19. 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

  20. The fusion reactor wall is getting hot. A challenge towards the future for numerical modelling (4). Chap. 4. What is really happening in the wall?

    International Nuclear Information System (INIS)

    Murata, Isao; Konno, Chikara

    2008-01-01

    In fusion plasmas, a lot of fast neutrons with a kinetic energy of 14 MeV are generated through D-T fusion reactions. These neutrons travel deep into the first wall and are absorbed in the blanket through nuclear reactions. In the present chapter, the authors discuss what happens in the blanket with the help of computerized simulation. (T.I.)

  1. 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.)

  2. 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.

  3. 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.

  4. 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.

  5. New methods for isolation of keratolytic bacteria inducing intractable hoof wall cavity (Gidoh) in a horse; double screening procedures of the horn powder agar-translucency test and horn zymography.

    Science.gov (United States)

    Kuwano, Atsutoshi; Niwa, Hidekazu; Arai, Katsuhiko

    2017-01-01

    To establish a new system to isolate keratolytic bacteria from the hoof wall cavity ( Gidoh ) of a racehorse, we invented the horn powder agar-translucency (HoPAT) test and horn zymography (HZ). Using routine bacteriological techniques and these methods, we isolated five strains of keratolytic soil bacteria, which were then identified by means of 16S ribosomal RNA (rRNA) gene sequencing analysis. The findings from the study on the horse suggested that Brevibacterium luteolum played the main role in the local fragility of the hoof, eventually forming a Gidoh in coordination with four other strains of keratolytic bacteria. The double screening procedures of the HoPAT test and HZ were useful and easy techniques for isolating the keratolytic bacteria from the horn lesions.

  6. Analysis of loss of electrical power 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.

    1994-08-01

    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 is similar to that reported in CFFTP-G--9355, Volume 4 except that a much larger decay power transient is used. Also, the pressurizer heaters are turned off following the loss of electrical power. This analysis is 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 of heatup of the in-core components. A description of the important aspects of the transient thermalhydraulic behaviour including coolant temperatures, circuit and sector flows, circuit pressure, pressurizer level and steam bleed flow, and first wall and blanket temperatures are provided. (author). 8 refs., 2 tabs., 26 figs

  7. A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Birch, M.; Brocklehurst, J.E.

    1987-12-01

    The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)

  8. Research on the wetted first wall concept for future laser fusion reactors. Final report No. 1, October 1, 1974--January 31, 1976

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Munir, Z.A.

    1976-01-01

    Research is in progress to determine the feasibility of the wetted first wall concept for a future laser fusion reactor. The basic idea involves the use of a thin coating of lithium on the inner wall of the laser fusion containment vessel to protect it from the micro-explosion blast debris. This report contains a review of the available information on contact angles and wettability of alkali metals on various metal substrates as well as a review of literature on thin falling liquid films. A proposed experiment to measure the contact angles of lithium on stainless steel and niobium is described. The requirements for a second experiment to measure certain key characteristics of thin falling films are also included

  9. Summary report of the IAEA advisory group meeting on nuclear data for neutron multiplication in fusion-reactor first-wall and blanket materials

    International Nuclear Information System (INIS)

    Muir, D.W.; Pashchenko, A.B.

    1992-09-01

    The present Report contains the Summary of the IAEA Advisory Group Meeting on Nuclear Data for Neutron Multiplication in Fusion-Reactor First-Wall and Blanket Materials, which was hosted by the Southwest Institute of Nuclear physics and Chemistry (SWINPC) at Chengdu, China and held from 19-21 November 1990. This AGM was organized by the IAEA Nuclear Data Section (NDS), with the cooperation and assistance of local organizers at the SWINPC. The papers which the participants prepared for and presented at the meeting will be published as an INDC report. (author)

  10. Simulations of fusion chamber dynamics and first wall response in a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR)

    Energy Technology Data Exchange (ETDEWEB)

    Qi, J.M., E-mail: qjm06@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Wang, Z., E-mail: wangz_es@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Chu, Y.Y., E-mail: chuyanyun@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Li, Z.H., E-mail: lee_march@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China)

    2016-03-15

    Highlights: • Z-FFR utilizes DT neutrons to drive a sub-critical fission blanket to produce energy. • A metal shell and Ar gas are employed in the fusion chamber for shock mitigation. • Massive materials can effectively mitigate the thermal heats on the chamber wall. • The W-coated Zr-alloy first wall exhibits good viability as a long-lived component. - Abstract: In a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR), the fusion target will produce enormous energy of ∼1.5 GJ per pulse at a frequency of 0.1 Hz. Almost 20% of the fusion energy yield, approximately 300 MJ, is released in forms of pulsed X-rays. To prevent the first wall from fatal damages by the intense X-rays, a thin spherical metal shell and rare Ar buffer gas are introduced to mitigate the transient X-ray bursts. Radiation hydrodynamics in the fusion chamber were investigated by MULTI-1D simulations, and the corresponding thermal and mechanical loads on the first wall were also obtained. The simulations indicated that by optimizing the design parameters of the metal shell and Ar buffer gas, peak power flux of the thermal heats on the first wall could be mitigated to less than 10{sup 4} W/cm{sup 2} within a time scale of several milliseconds, while peak overpressures of the mechanical loads varying from 0.6 to 0.7 MPa. In addition, the thermomechanical response in a W–coated Zr-alloy first wall was performed by FWDR1D calculations using the derived thermal and mechanical loads as inputs. The temperature and stress fields were analyzed, and the corresponding elastic strains were conducted for primary lifetime estimations by using the Coffin–Manson relationships of both W and Zr-alloy. It was shown that the maximum temperature rises and stresses in the first wall were less than 50 K and 130 MPa respectively, and lifetime of the first wall would be in excess of 10{sup 9} cycles. The chamber exhibits good viability as a long-lived component to sustain the Z-FFR conceptual

  11. 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

  12. accelerating cavity

    CERN Multimedia

    On the inside of the cavity there is a layer of niobium. Operating at 4.2 degrees above absolute zero, the niobium is superconducting and carries an accelerating field of 6 million volts per metre with negligible losses. Each cavity has a surface of 6 m2. The niobium layer is only 1.2 microns thick, ten times thinner than a hair. Such a large area had never been coated to such a high accuracy. A speck of dust could ruin the performance of the whole cavity so the work had to be done in an extremely clean environment.

  13. 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

  14. 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)

  15. 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

  16. radiofrequency cavity

    CERN Multimedia

    1988-01-01

    The pulse of a particle accelerator. 128 of these radio frequency cavities were positioned around CERN's 27-kilometre LEP ring to accelerate electrons and positrons. The acceleration was produced by microwave electric oscillations at 352 MHz. The electrons and positrons were grouped into bunches, like beads on a string, and the copper sphere at the top stored the microwave energy between the passage of individual bunches. This made for valuable energy savings as it reduced the heat generated in the cavity.

  17. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  18. 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

  19. Several loadings and stresses of first wall of SiC with metal liner on conceptual design of moving ring reactor 'KARIN-1'

    International Nuclear Information System (INIS)

    Nishikawa, Masahiro; Tachibana, Eizaburo; Watanabe, Kenji; Fujiie, Yoichi.

    1983-01-01

    On conceptual design of moving ring reactor ''KARIN-I'' (Output: 1850 MWe), the first wall of SiC with metal liner is considered by reason that SiC ceramics has specific features of excellent radiation damage resistance in fast neutron spectra and a very low residual radioactivity, and that the thin metal liner has good compatibility with liquid lithium and good vaccum-tight, however, a extent electromagnetic interaction. The electromagnetic force applied on the metal liner and several pressure losses of liquid lithum flow are estimated, and these forces correspond to the fluid mechanical loading on SiC first wall. Thermal loading by neutron flux is calculated on the first wall to obtain temperature distributions along the flow direction and toward the wall thickness. At the outlet of the burning section, the surface temperature of SiC rises to the value of 825 0 C on plasma side and on the metal liner, it rises to the value of 540 0 C. Finally, the stress analysis is performed. The thermal stress is about one order larger than the stress induced by the fluid mechanical loading. At the inlet of the burning section, the average tensile stress of 22.4kg/mm 2 is induced on the outer side of SiC wall, and on the inner side, the average compressive stress of -26.1kg/mm 2 is induced. At the outlet of the burning section, the tensile stress is found to oscillate between 25.5kg/mm 2 and 27.3kg/mm 2 on the outer side of SiC wall by frequency of 1 Hz, and on the inner side, the compressive stress also oscillates between -21.6kg/mm 2 and -29.0kg/mm 2 by the same frequency. These stresses are within the value of fracture stress, (72.5kg/mm 2 ). Difficult residual problems on the first wall are also discussed. (author)

  20. Thermo-mechanical design and structural analysis of the first wall for ARIES-III, A 1000 MWeD-3He power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.; Blanchard, J.P.; Mogahed, E.A.

    1992-01-01

    This paper reports on ARIES III, a conceptual design study of a 1000 MWe D- 3 He tokamak fusion power reactor in which most of the energy comes from charged particle transport, bremsstrahlung and synchrotron radiation, and only a small fraction (∼ 4%) comes form neutrons. This form of energy is deposited as surface heating on the chamber first wall (FW) and divertor elements, while the neutron energy is deposited as bulk nuclear heating within the shield. Since this reactor does not use tritium, there is no breeding blanket. Instead a shield is provided to protect the magnets from neutrons. The Fw is very unique in a D- 3 He reactor, it must be capable of absorbing the high surface heat in a mode suitable for efficient power cycle conversion, it must be able to reflect synchrotron radiation, and it must be able to withstand high current plasma disruptions. The FW is made of a low activation ferritic steel (MHT-9) and is cooled with an organic coolant (HB-40) at a pressure of 2 MPa. The FW has a coating of 0.01 cm tungsten on the MHT-9, followed by 0.15 cm of Be on the plasma side. This is needed for synchrotron radiation reflection and as a melt layer to guard against the thermal effects of a plasma disruption

  1. 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....

  2. Cold-walled UHV/CVD batch reactor for the growth of Si1_x/Gex layers

    DEFF Research Database (Denmark)

    Thomsen, Erik Vilain; Christensen, Carsten; Andersen, C.R.

    1997-01-01

    A novel cold-walled, lamp-heated, ultrahigh vacuum chemical vapor deposition (UHV/CVD) batch system for the growth of SiGe layers is presented. This system combines the batch capability of the standard UHV/CVD furnace with the temperature processing available in rapid thermal processing (Rm...

  3. Plasma non-uniformity in a symmetric radiofrequency capacitively-coupled reactor with dielectric side-wall: a two dimensional particle-in-cell/Monte Carlo collision simulation

    Science.gov (United States)

    Liu, Yue; Booth, Jean-Paul; Chabert, Pascal

    2018-02-01

    A Cartesian-coordinate two-dimensional electrostatic particle-in-cell/Monte Carlo collision (PIC/MCC) plasma simulation code is presented, including a new treatment of charge balance at dielectric boundaries. It is used to simulate an Ar plasma in a symmetric radiofrequency capacitively-coupled parallel-plate reactor with a thick (3.5 cm) dielectric side-wall. The reactor size (12 cm electrode width, 2.5 cm electrode spacing) and frequency (15 MHz) are such that electromagnetic effects can be ignored. The dielectric side-wall effectively shields the plasma from the enhanced electric field at the powered-grounded electrode junction, which has previously been shown to produce locally enhanced plasma density (Dalvie et al 1993 Appl. Phys. Lett. 62 3207-9 Overzet and Hopkins 1993 Appl. Phys. Lett. 63 2484-6 Boeuf and Pitchford 1995 Phys. Rev. E 51 1376-90). Nevertheless, enhanced electron heating is observed in a region adjacent to the dielectric boundary, leading to maxima in ionization rate, plasma density and ion flux to the electrodes in this region, and not at the reactor centre as would otherwise be expected. The axially-integrated electron power deposition peaks closer to the dielectric edge than the electron density. The electron heating components are derived from the PIC/MCC simulations and show that this enhanced electron heating results from increased Ohmic heating in the axial direction as the electron density decreases towards the side-wall. We investigated the validity of different analytical formulas to estimate the Ohmic heating by comparing them to the PIC results. The widespread assumption that a time-averaged momentum transfer frequency, v m , can be used to estimate the momentum change can cause large errors, since it neglects both phase and amplitude information. Furthermore, the classical relationship between the total electron current and the electric field must be used with caution, particularly close to the dielectric edge where the (neglected

  4. 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.)

  5. 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)

  6. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  7. Phase instability and toughness change during high temperature exposure of various steels for the first wall structural materials of a fusion reactor

    International Nuclear Information System (INIS)

    Miyahara, K.; Shimoide, Y.

    1995-01-01

    The objective of the present research is to clarify the phase instability, particularly, the precipitation behavior of carbide and nitride during the long term aging in the non-irradiation state of the materials proposed for the first wall structural component of fusion reactors, such as a type 316 austenitic steel, its modified steels, ferritic heat resisting steels and reduced radio-activation materials. The effect of the precipitation behavior on the toughness is also investigated. It is noticed that the toughness was much deteriorated by the formation of large amounts of coarse carbides within grains and on grain boundaries during 2.88x10 4 ks (8000 h) aging at 873 K and that intergranular fracture occurred by the impact test at room temperature even in the type 316 steel. (orig.)

  8. 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

  9. 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.

  10. 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

  11. Modelling of local steam condensation on walls in presence of noncondensable gases. Application to a local calculation in reactor containment using the multidimensional geyser code

    International Nuclear Information System (INIS)

    Benet, L.V.; Caroli, C.; Cornet, P.; Coulon, N.; Magnaud, J.P.

    1995-01-01

    The frame of this paper is the study of severe accidents of pressurized water reactors (PWR). The need for containment modelling, and in particular for hydrogen risk study, was reinforced in France after 1990, with the requirement of taking into account severe accidents in design of future plants. This new need of assessing the transient local hydrogen concentration incited us to develop the multidimensional code GEYSER for containment analysis. This code, developed by the Department of Mechanics and Technology of the French Atomic Energy Commission, is presented here with a detailed example of calculation. We describe the mixture, whose constituents are noncondensable gases (air or air and hydrogen) and water vapor liquid, by a compressible homogeneous diphasic model. The wall condensation is based on the Chilton-Colburn formulation and heat mass transfer analogy. We present a transient two-dimensional axisymmetric calculation of a simplified accidental sequence during one hour. The calculation in the large volume of a reactor containment consists of an injection of vapor, first important then moderate, followed by an injection of hydrogen. (authors). 8 refs., 4 figs., 4 tabs

  12. Validation of nonlinear FEA models of a thin-walled elbow under extreme loading conditions for Sodium-cooled Fast Reactors

    International Nuclear Information System (INIS)

    Watakabe, Tomoyoshi; Wakai, Takashi; Jin, Chuanrong; Usui, Yoshiya; Sakai, Shinkichi; Ooshika, Junji; Tsukimori, Kazuyuki

    2015-01-01

    For the purpose of confirming failure modes and safety margin, some studies on the ultimate strength of thin-walled piping components for Sodium-cooled Fast Reactors (SFRs) under extreme loading conditions such as large earthquakes have been reported these several years. Nonlinear finite element analysis has been applied in these studies to simulate buckling and yielding with large deformation, whose accuracy is dependent on the element type, the mesh size, the elasto-plastic model and so on. It is important to check the validation of a finite element model for nonlinear analysis especially under extreme loading conditions. This paper presents static and dynamic analyses of a thin-walled elbow with large deformation under large seismic loading, and discusses the validation of the FEA models comparing with experimental results. The finite element analysis models in this study are generated by shell elements for a stainless steel pipe elbow of diameter-to-thickness ratio 59:1 similar to the main pipe of SFRs, which is used for shaking table tests. At first, a static analysis is carried out for an in-plane monotonic bending test, in order to confirm that the shell element is appropriate to the large deformation analysis and the material parameters are proper for the strain level in the experiments. And then, a dynamic in-plane bending test with the maximum acceleration of 11.7G is simulated by the nonlinear FEA with stiffness-proportional damping. The influence of mesh sizes on results is investigated, to determine proper mesh sizes and reduce the computational cost. Finally, comparing the results of the FEM analyses with those of experiments, it is concluded that the appropriately generated FEA models are effective and give accurate results for nonlinear analyses of the thin-walled elbow under large seismic loading. (author)

  13. 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....

  14. Report of second meeting on the interaction of plasma and the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Yamashina, Toshiro; Watanabe, Kuniaki; Mori, Mamoru; Tominaga, Goro; Kinbara, Akira.

    1979-10-01

    This report presents various problems on the interaction between plasma and materials. The first half of this report is the reports of international meetings. First topical meeting on fusion reactor materials, IEA-Textor workshop on surface measurements, and sixth international vacuum metallurgy conference on special melting and metallurgical coatings are summarized. The other half of the report is described on the present and future plans of the analysis of material surfaces which are carried out at the laboratories in Japan. The last part of the report introduces the TEXTOR international cooperative study project. (Kato, T.)

  15. Operation method for wall surface of pressure suppression chamber of reactor container and floating scaffold used for the method

    International Nuclear Information System (INIS)

    Matsuzaki, Tetsuo; Kounomaru, Toshimi; Saito, Koichi.

    1996-01-01

    A floating scaffold is provisionally disposed in adjacent with the wall surface of pool water of a pressure suppression chamber while being floated on the surface of the pool water before the drainage of the pool water from the pressure vessel. The floating scaffold has guide rollers sandwiching a bent tube of an existent facility so that the horizontal movement is restrained, and is movable only in a vertical direction depending on the change of water level of the pool water. In addition, a handrail for preventing dropping, and a provisional illumination light are disposed. When pool water in the pressure suppression chamber is drained, the water level of the pool water is lowered in accordance with the amount of drained water. The floating scaffold floating on the water surface is lowered while being guided by the bent tube, and the operation position is lowered. An operator riding on the floating scaffold inspects the wall surfaces of the pressure chamber and conducts optional repair and painting. (I.N.)

  16. Research and tests of steel-concrete-steel sandwich composite shear wall in reactor containment of HTR-PM

    International Nuclear Information System (INIS)

    Sun Yunlun; Huang Wen; Zhang Ran; Zhang Pei; Tian Chunyu

    2014-01-01

    By quasi-static test of 8 specimens of steel-concrete-steel sandwich composite shear wall, the bearing capacity, hysteretic behavior, failure mode of the specimens was studied. So was the effect of the shear-span ratios, steel ratios and spacing of studs on the properties of the specimens. The failure patterns of all specimens with different shear-span ratios between 1.0 and 1.5 were compression-bending failure. The hysteretic curves of all specimens were relatively plump, which validated the well deformability and energy dissipation capacity of the specimens. When shear-span ratio less than 1.5, the shear property of the steel plate was well played, and so was the deformability of the specimens. The bigger the steel ratio was, the better the lateral resistance capacity and the deformability was. Among the spacing of studs in the test, the spacing of studs had no significant effect on the bearing capacity, deformability and ductility of the specimens. Based on the principle of superposition an advised formula for the compression-bending capacity of the shear wall was proposed, which fitted well with the test result and had a proper safety margin. (author)

  17. Finite element elasto-plastic analysis of thin walled structures of reinforced concrete as applied to reactor facilities

    International Nuclear Information System (INIS)

    Fujita, F.; Tsuboi, Y.

    1981-01-01

    The authors developed a new program of elasto-plastic analysis of reinforced concrete shells, in which the simplest model of shell element and an orthotropic constitutive relation are adopted, and verified its validity with reference to the results of model experiments of containers and box-wall structures with various loading conditions. For the two-dimensional stress-strain relationship of concrete, an orthotropic nonlinear formula proposed by one of the authors was adopted. For concrete, the octahedral shear failure and tension cut-off criteria were also imposed. The Kirchhoff-Love's assumptions were assumed to be valid for the whole range of the analysis and the layered approach of elasto-plastic stiffness evaluation. Derivation of the shell element is outlined with examination of its accuracy in elastic range and the assumption of elasto-plastic material property and the procedure of nonlinear analysis are described. As examples, the method is applied to the analysis of a cylindrical container and a box-wall structure. Comparison of the computed results with the corresponding experimental data indicates the applicability of the proposed method. (orig./HP)

  18. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    International Nuclear Information System (INIS)

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods

  19. An intercomparison of HO2 measurements by fluorescence assay by gas expansion and cavity ring-down spectroscopy within HIRAC (Highly Instrumented Reactor for Atmospheric Chemistry

    Directory of Open Access Journals (Sweden)

    L. Onel

    2017-12-01

    Full Text Available The HO2 radical was monitored simultaneously using two independent techniques in the Leeds HIRAC (Highly Instrumented Reactor for Atmospheric Chemistry atmospheric simulation chamber at room temperature and total pressures of 150 and 1000 mbar of synthetic air. In the first method, HO2 was measured indirectly following sampling through a pinhole expansion to 3 mbar when sampling from 1000 mbar and to 1 mbar when sampling from 150 mbar. Subsequent addition of NO converted it to OH, which was detected via laser-induced fluorescence spectroscopy using the FAGE (fluorescence assay by gas expansion technique. The FAGE method is used widely to measure HO2 concentrations in the field and was calibrated using the 185 nm photolysis of water vapour in synthetic air with a limit of detection at 1000 mbar of 1.6 × 106 molecule cm−3 for an averaging time of 30 s. In the second method, HO2 was measured directly and absolutely without the need for calibration using cavity ring-down spectroscopy (CRDS, with the optical path across the entire ∼ 1.4 m width of the chamber, with excitation of the first O-H overtone at 1506.43 nm using a diode laser and with a sensitivity determined from Allan deviation plots of 3.0 × 108 and 1.5 × 109 molecule cm−3 at 150 and 1000 mbar respectively, for an averaging period of 30 s. HO2 was generated in HIRAC by the photolysis of Cl2 using black lamps in the presence of methanol in synthetic air and was monitored by FAGE and CRDS for ∼ 5–10 min periods with the lamps on and also during the HO2 decay after the lamps were switched off. At 1000 mbar total pressure the correlation plot of [HO2]FAGE versus [HO2]CRDS gave an average gradient of 0.84 ± 0.08 for HO2 concentrations in the range ∼ 4–100 × 109 molecule cm−3, while at 150 mbar total pressure the corresponding gradient was 0.90 ± 0.12 on average for HO2 concentrations in the range

  20. 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

  1. 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

  2. Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

    International Nuclear Information System (INIS)

    Sato, Satoshi; Maegawa, Toshio; Yoshimatsu, Kenji; Sato, Koichi; Nonaka, Akira; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

    2011-01-01

    A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

  3. Economic aspect comparison between steel plate reinforced concrete and reinforced concrete technique in reactor containment wall construction

    International Nuclear Information System (INIS)

    Yuliastuti; Sriyana

    2008-01-01

    Construction costs of nuclear power plant were high due to the construction delays, regulatory delays, redesign requirement, and difficulties in construction management. Based on US DOE (United States Department of Energy) study in 2004, there were thirteen advanced construction technologies which were potential to reduce the construction time of nuclear power plant. Among these technologies was the application of steel-plate reinforced concrete (SC) on reactor containment construction. The conventional reinforced concrete (RC) technique were built in place and require more time to remove framework since the external form is temporary. Meanwhile, the SC technique offered a more efficient way to placing concrete by using a permanent external form made of steel. The objective of this study was to calculate construction duration and economic comparison between RC and SC technique. The result of this study showed that SC technique could reduce the construction time by 60% and 29,7% cost reduced compare to the RC technique. (author)

  4. Recycling, inventory and permeation of hydrogen isotopes and helium in the first wall of a thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Gervasini, G.; Reiter, F.

    1989-01-01

    The work was divided into three parts. The first part, which is theoretical, examines the behaviour of hydrogen in metals. After an introduction on the presence of hydrogen isotopes in fusion reactors, the main phenomena connected with hydrogen-metal interaction are summarised: solubility, diffusivity and trapping in material defects. The metal temperature is highlighted as the main parameter in the description of the phenomena. The second part of the work, also theoretical, concerns the interaction between helium and metals. We have tried as much as possible to show analogies and differences in the comparisons of the behaviour of hydrogen. The main types of damage caused by helium in metallic structures, which are the most important consequence of helium-metal interaction, were summarised. The characteristics of helium were treated in greater depth than those of hydrogen, because the latter are very well known. Also, there is a vast literature on the hydrogen-metal interaction. In the third and last part of the work a model was identified which allows the simulation of the evolution of a system formed from a metal in which hydrogen and helium isotopes have been introduced. A system of algebraic-differential equations was used to study the temporal evolution of the concentrations, the recycling, the inventory and the permeation of tritium and helium considering that these atoms diffuse in the metallic lattice and remain trapped in the vacancies created inside the metal by the bombardment of the neutrons from the fusion reactions. For the numerical simulation a series of data intended to represent the situation inside a thermonuclear reactor as precisely as possible were used for the numerical simulation. Analysis of the system was preceded by the analytical resolution of the steady state equations so that they could be compared with the simulation results

  5. 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.)

  6. Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Simpson, R.B.

    1995-01-01

    Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation

  7. 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

  8. First wall/blanket/shield design and power conversion for the ARIES-IV tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Conn, R.W.; Najmabadi, F.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10 MPa base pressure. The coolant flows poloidally in two loops, one inboard and one outboard. The coolant channels are circular tubes that form shells and are placed between two purge plates; the space between two adjacent tubes and the plate is purge gas flow area. The solid breeder is Li 2 O, and Be is used as neutron multiplier to ensure adequate TBR. Beryllium and Li 2 O are placed in between the adjacent tube shells. A computer code was developed to perform and optimize thermal-hydraulic design. Minimization of blanket thickness and the amount of Be, and the maximization of breeder zone thickness were done by iteration with neutronics. The gross thermal efficiency is 49%. The cost of electricity is 68 mills/kWh. The use of low activation SiC composite as the structural material, Li 2 O as the solid breeder, and avoidance of tungsten in the divertor has resulted in a good safety performance, and LSA rating of 1. Overall, SiC/He/Li 2 O ARIES-IV design is expected to have attractive economic and safety advantages

  9. Finite element analysis of the shock waves induced in the liquid wall of a pellet fusion reactor

    International Nuclear Information System (INIS)

    Miya, K.; Iizuka, T.; Silverman, J.

    1985-01-01

    A shock wave induced in liquid metal is analyzed numerically by application of the finite element method. Since the governing equations of motion of the fluid are nonlinear, an incremental method is combined with the finite element method to obtain a convergent solution of the shock wave without an interaction technique. To demonstrate the validity of the method developed, shock wave problems in an inertial confinement spherical reactor with a liquid lithium ''waterfall'' are solved for two cases of surface heating due to soft x-ray absorption and bulk heating due to 14-MeV neutron absorption. The solution is based on a combination of the conservation equations for mass, energy, and momentum along with the following equation of state for liquid metals: p = P /sub b/ ((/rho///rho/ 0 ) /sup n/ - 1). Numerical results show that peak pressure induced in the liquid lithium is very high even for a comparatively small energy release E /sub TAU/ = 100 MJ/microexplosion of a pellet. Dynamic stress induced in a 5-cm-thick stainless steel pressure vessel is 1.14 x 10 3 MPa for the surface heating. The results show that the dynamic stress induced by bulk heating is superimposed on that due to surface heating within the same period. Two appropriate ways to reduce the high stress are application of two-phase flow of liquid lithium or an increase in the thickness of the pressure vessel

  10. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

    1992-09-01

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  11. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  12. 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.)

  13. 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

  14. 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.)

  15. Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow

    International Nuclear Information System (INIS)

    Whaley, R.L.; Sanders, J.P.

    1976-09-01

    A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection

  16. Results of charpy V-notch impact testing of structural steel specimens irradiated at ∼30 degrees C to 1 x 1016 neutrons/cm2 in a commercial reactor cavity

    International Nuclear Information System (INIS)

    Iskander, S.K.; Stoller, R.E.

    1997-04-01

    A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at ∼ 30 degrees C (∼ 85 degrees F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10 16 neutrons/cm 2 (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was ∼ 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of ∼ 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications

  17. Construction and characterization of a new high current ion source for research of impact of hydrogen irradiation on wall materials for use in nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Arredondo Parra, Rodrigo; Neu, Rudolf [Max Planck Institute for Plasma Physics, Garching (Germany); Technische Universitaet Muenchen, Garching (Germany); Oberkofler, Martin; Schmid, Klaus; Weghorn, Arno [Max Planck Institute for Plasma Physics, Garching (Germany)

    2016-07-01

    The HSQ (HochStromQuelle) is a high current DuoPIGatron type ion source used for research in surface properties of wall materials for nuclear fusion reactors. The existing HSQ-I will be replaced by the conceptually identical HSQ-II, currently under construction. Varying the acceleration potential and optimizing gas inflow and beam focusing grid voltage, ion currents before the deflecting magnet between 10 and 875 μA were reached for acceleration voltages of 0.7 to 8 kV. The ion beam footprint will be characterized, and ion optics will be installed before and after the deflecting magnet, capable of bending 10 keV Ar. A monoenergetic beam of a single species (e.g. D{sub 3}{sup +}) will finally be used for irradiation of samples in the separate implantation chamber at a base pressure of 10{sup -8} mbar. The energy of the impinging particles ranges from 200 eV/D to several keV/D. Fluxes of 10{sup 15} D/cm{sup 2}/s to the target are expected. The temperature of the sample is varied via electron impact heating and the sample weight can be assessed in situ by means of a magnetic suspension balance.

  18. 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.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  20. 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

  1. 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

  2. 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.

  3. 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

  4. 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)

  5. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  6. 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

  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

  8. Characterization of graded iron / tungsten layers for the first wall of fusion reactors; Charakterisierung gradierter Eisen/Wolfram-Schichten fuer die erste Wand von Fusionsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Heuer, Simon

    2017-07-01

    The nuclear fusion has great potential to enable a CO{sub 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. [German] Die Kernfusion besitzt grosses Potenzial eine CO{sub 2}-neutrale Energieversorgung zukuenftiger Generationen zu ermoeglichen. Dabei stellt die technische Nutzbarmachung dieser

  9. 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

  10. Using NJOY99 and MCNP4B2 to Estimate the Radiation Damage Displacements per Atom per Second in Steel Within the Boiling Water Reactor Core Shroud and Vessel Wall from Reactor-Grade Mixed-Oxide/Uranium Oxide Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

    International Nuclear Information System (INIS)

    Vickers, Lisa

    2003-01-01

    The government of Mexico has expressed interest in utilizing the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 to 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons.There is concern that a core with a fraction of MOX fuel (i.e., increased 239 Pu wt%) would increase the radiation damage displacements per atom per second (dpa-s -1 ) in steel within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation damage within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor.The primary uniqueness of this paper is the computation of radiation damage (dpa-s -1 ) using NJOY99-processed cross sections for steel within the core shroud and vessel wall. Specifically, the unique radiation damage results are several orders of magnitude greater than results of previous works. In addition, the conclusion of this paper was that the addition of the maximum fraction of one-third MOX fuel to the LV1 BWR core did significantly increase the radiation damage in steel within the core shroud and vessel wall such that without mitigation of radiation damage by periodic thermal annealing or reduction in operating parameters such as neutron fluence, core temperature, and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor

  11. 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.

  12. 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.

  13. 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)

  14. 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

  15. Thermal Shock Experiment (TSEX): a ''proof-of-principle'' evaluation of the use of electron beam heating to simulate the thermal mechanical environment anticipated for the first wall of the Reference Theta-Pinch Reactor (RTPR)

    International Nuclear Information System (INIS)

    Armstrong, P.E.; Krakowski, R.A.

    1977-06-01

    The results of a ''proof-of-principle'' Thermal Shock Experiment (TSEX), designed to simulate the thermal mechanical response of insulator-metal composite first walls anticipated for pulsed high-density fusion reactors, are given. A programmable 10-kV, 1.0-A electron beam was used to pulse repeatedly (0.30-mm)Al 2 O 3 /(1.0-mm) Nb-1Zr composite samples 200 to 300 K, relative to a base-line temperature of 1000 K. The experimental goals of TSEX were established relative to the first-wall environment anticipated for the Reference Theta-Pinch Reactor (RTPR). A detailed description of the TSEX ''proof-of-principle'' apparatus, experimental procedure, and diagnostics is given. The results of extensive thermal analyses are given, which are used to estimate the thermal stresses generated. Although little or no control was exercised over the sample fabrication and thermal history, one sample experienced in excess of 800 thermal cycles of approximately 250 K at approximately 1000 K, and the results of optical and SEM examination of this specimen are presented. The resistance of this sample to macroscopic failure was truly impressive. Recommendations for the construction of an apparatus dedicated to extensive testing of first-wall composites are given on the basis of these ''proof-of-principle'' TSEX results

  16. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  17. 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.

  18. 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.

  19. 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

  20. 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...

  1. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  2. 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

  3. 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.

  4. Fusion: first wall problems

    International Nuclear Information System (INIS)

    Behrisch, R.

    1976-01-01

    Some of the relevant elementary atomic processes which are expected to be of significance to the first wall of a fusion reactor are reviewed. Up to the present, most investigations have been performed at relatively high ion energies, typically E greater than 5 keV, and even in this range the available data are very poor. If the plasma wall interaction takes place at energies of E greater than 1 keV the impurity introduction and first wall erosion which will take place predominantly by sputtering, will be large and may severely limit the burning time of the plasma. The wall bombardment and surface erosion will presumably not decrease substantially by introducing a divertor. The erosion can only be kept low if the energy of the bombarding ions and neutrals can be kept below the threshold for sputtering of 1 to 10 eV. 93 refs

  5. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  6. 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 ...

  7. US/Japan collaborative program on fusion reactor materials: Summary of the tenth DOE/JAERI Annex I technical progress meeting on neutron irradiation effects in first wall and blanket structural materials

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1989-01-01

    This meeting was held at Oak Ridge National Laboratory on March 17, 1989, to review the technical progress on the collaborative DOE/JAERI program on fusion reactor materials. The purpose of the program is to determine the effects of neutron irradiation on the mechanical behavior and dimensional stability of US and Japanese austenitic stainless steels. Phase I of the program focused on the effects of high concentrations of helium on the tensile, fatigue, and swelling properties of both US and Japanese alloys. In Phase II of the program, spectral and isotropic tailoring techniques are fully utilized to reproduce the helium:dpa ratio typical of the fusion environment. The Phase II program hinges on a restart of the High Flux Isotope Reactor by mid-1989. Eight target position capsules and two RB* position capsules have been assembled. The target capsule experiments will address issues relating to the performance of austenitic steels at high damage levels including an assessment of the performance of a variety of weld materials. The RB* capsules will provide a unique and important set of data on the behavior of austenitic steels irradiated under conditions which reproduce the damage rate, dose, temperature, and helium generation rate expected in the first wall and blanket structure of the International Thermonuclear Experimental Reactor

  8. 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.

  9. 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...

  10. Conditions of vacuum physics for selection of the material of first wall and diaphragm of the demonstration thermonuclear reactor-tokamak (T-20)

    International Nuclear Information System (INIS)

    Gusev, V.M.; Guseva, M.I.; Gervids, V.I.; Kogan, V.I.; Martynenko, Yu.V.; Mirnov, S.V.

    A model is given for plasma interaction with the wall and the introduction of contaminants. The model was characterized by two kinds of uncertainty. First, the uncertain behavior of the contaminants, and second, the uncertainty of boundary conditions. Some of the conclusions from the study are described

  11. The LHC superconducting cavities

    CERN Document Server

    Boussard, Daniel; Häbel, E; Kindermann, H P; Losito, R; Marque, S; Rödel, V; Stirbet, M

    1999-01-01

    The LHC RF system, which must handle high intensity (0.5 A d.c.) beams, makes use of superconducting single-cell cavities, best suited to minimizing the effects of periodic transient beam loading. There will be eight cavities per beam, each capable of delivering 2 MV (5 MV/m accelerating field) at 400 MHz. The cavities themselves are now being manufactured by industry, using niobium-on-copper technology which gives full satisfaction at LEP. A cavity unit includes a helium tank (4.5 K operating temperature) built around a cavity cell, RF and HOM couplers and a mechanical tuner, all housed in a modular cryostat. Four-unit modules are ultimately foreseen for the LHC (two per beam), while at present a prototype version with two complete units is being extensively tested. In addition to a detailed description of the cavity and its ancillary equipment, the first test results of the prototype will be reported.

  12. LEP copper accelerating cavities

    CERN Multimedia

    Laurent Guiraud

    1999-01-01

    These copper cavities were used to generate the radio frequency electric field that was used to accelerate electrons and positrons around the 27-km Large Electron-Positron (LEP) collider at CERN, which ran from 1989 to 2000. The copper cavities were gradually replaced from 1996 with new superconducting cavities allowing the collision energy to rise from 90 GeV to 200 GeV by mid-1999.

  13. 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.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  15. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  16. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  17. 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

  18. 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

  19. 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

  20. Development of an application-oriented multi-frequency eddy current procedure for the outer reactor vessel- and store vessel wall of the SNR-300

    International Nuclear Information System (INIS)

    Hoeft, E.

    1991-08-01

    The following companies participated in the development of the application oriented multi-frequency eddy current procedure for the outer reactor vessel- and store vesselwall of the SNR-300: Interatom GmbH (coordinator), MAN-Energie GmbH (ME, subcontractor), Fraunhofer Institut IzfP, own promotion project). The precisely defined work packages of the participating companies Interatom and IzfP were supported by the Federal Minister for Research and Technology in separate promotion project. The present report comprises the work performed at Interatom and ME for developing the manipulator and the subsystems. The development aim was reached largely. Manufactoring of the manipulator with all necessary peripherical equipments was finished and accepted in partial function tests at the manufacturer. Tests at the Interatom teststand however with the fully mounted systems at ambient- and reactor temperature could not be done within the appropriated timeschedule and finance frame. (orig.) [de

  1. Study on restriction method for end-wall boundary layer thickness in axial helium gas compressor for gas turbine high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kunitomi, Kazuhiko; Inagaki, Yoshiyuki

    2009-01-01

    Aerodynamic performance test was carried out using a 1/3 scale, 4-stage model of the helium gas compressor to investigate an effect of end-wall over-camber to prevent decrease of axial velocity in the end-wall boundary layer. The model compressor consists of a rotor, 500 mm in diameter, which is driven by an electric motor at a rotational speed of 10800 rpm. The rotor blade span of the first stage is 34 mm. The test was carried out under the condition that the helium gas pressure of 0.88 MPa, temperature of 30degC, and mass flow rate of 12.47 kg/s at the inlet. A 3-dimensional aerodynamic code, which was verified using the test data, showed that axial velocity was lowered by using a blade which increased the inlet blade angle around the end-wall region of the casing side in comparison with that using the original design blade, because the inlet flow angle mismatched with the inlet blade angle of the rotor blade, as opposed to the prediction by a conventional air compressor design method. The overall adiabatic efficiency of the full scale 20-stage helium gas compressor was predicted 89.7% from the Reynolds number dependency of the test data by using the original design blade. (author)

  2. The influence of spherical cavity surface charge distribution on the sequence of partial discharge events

    International Nuclear Information System (INIS)

    Illias, Hazlee A; Chen, George; Lewin, Paul L

    2011-01-01

    In this work, a model representing partial discharge (PD) behaviour of a spherical cavity within a homogeneous dielectric material has been developed to study the influence of cavity surface charge distribution on the electric field distribution in both the cavity and the material itself. The charge accumulation on the cavity surface after a PD event and charge movement along the cavity wall under the influence of electric field magnitude and direction has been found to affect the electric field distribution in the whole cavity and in the material. This in turn affects the likelihood of any subsequent PD activity in the cavity and the whole sequence of PD events. The model parameters influencing cavity surface charge distribution can be readily identified; they are the cavity surface conductivity, the inception field and the extinction field. Comparison of measurement and simulation results has been undertaken to validate the model.

  3. The influence of spherical cavity surface charge distribution on the sequence of partial discharge events

    Energy Technology Data Exchange (ETDEWEB)

    Illias, Hazlee A [Department of Electrical Engineering, Faculty of Engineering, University of Malaya, 50603 Kuala Lumpur (Malaysia); Chen, George; Lewin, Paul L, E-mail: h.illias@um.edu.my [Tony Davies High Voltage Laboratory, School of Electronics and Computer Science, University of Southampton, Southampton, SO17 1BJ (United Kingdom)

    2011-06-22

    In this work, a model representing partial discharge (PD) behaviour of a spherical cavity within a homogeneous dielectric material has been developed to study the influence of cavity surface charge distribution on the electric field distribution in both the cavity and the material itself. The charge accumulation on the cavity surface after a PD event and charge movement along the cavity wall under the influence of electric field magnitude and direction has been found to affect the electric field distribution in the whole cavity and in the material. This in turn affects the likelihood of any subsequent PD activity in the cavity and the whole sequence of PD events. The model parameters influencing cavity surface charge distribution can be readily identified; they are the cavity surface conductivity, the inception field and the extinction field. Comparison of measurement and simulation results has been undertaken to validate the model.

  4. Cavity design programs

    International Nuclear Information System (INIS)

    Nelson, E.M.

    1996-01-01

    Numerous computer programs are available to help accelerator physicists and engineers model and design accelerator cavities and other microwave components. This article discusses the problems these programs solve and the principles upon which these programs are based. Some examples of how these programs are used in the design of accelerator cavities are also given

  5. Cavity quantum electrodynamics

    International Nuclear Information System (INIS)

    Walther, Herbert; Varcoe, Benjamin T H; Englert, Berthold-Georg; Becker, Thomas

    2006-01-01

    This paper reviews the work on cavity quantum electrodynamics of free atoms. In recent years, cavity experiments have also been conducted on a variety of solid-state systems resulting in many interesting applications, of which microlasers, photon bandgap structures and quantum dot structures in cavities are outstanding examples. Although these phenomena and systems are very interesting, discussion is limited here to free atoms and mostly single atoms because these systems exhibit clean quantum phenomena and are not disturbed by a variety of other effects. At the centre of our review is the work on the one-atom maser, but we also give a survey of the entire field, using free atoms in order to show the large variety of problems dealt with. The cavity interaction can be separated into two main regimes: the weak coupling in cavity or cavity-like structures with low quality factors Q and the strong coupling when high-Q cavities are involved. The weak coupling leads to modification of spontaneous transitions and level shifts, whereas the strong coupling enables one to observe a periodic exchange of photons between atoms and the radiation field. In this case, atoms and photons are entangled, this being the basis for a variety of phenomena observed, some of them leading to interesting applications in quantum information processing. The cavity experiments with free atoms reached a new domain with the advent of experiments in the visible spectral region. A review on recent achievements in this area is also given

  6. Formation of coronal cavities

    International Nuclear Information System (INIS)

    An, C.H.; Suess, S.T.; Tandberg-Hanssen, E.; Steinolfson, R.S.

    1986-01-01

    A theoretical study of the formation of a coronal cavity and its relation to a quiescent prominence is presented. It is argued that the formation of a cavity is initiated by the condensation of plasma which is trapped by the coronal magnetic field in a closed streamer and which then flows down to the chromosphere along the field lines due to lack of stable magnetic support against gravity. The existence of a coronal cavity depends on the coronal magnetic field strength; with low strength, the plasma density is not high enough for condensation to occur. Furthermore, we suggest that prominence and cavity material is supplied from the chromospheric level. Whether a coronal cavity and a prominence coexist depends on the magnetic field configuration; a prominence requires stable magnetic support

  7. 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.)

  8. SPS RF Accelerating Cavity

    CERN Multimedia

    1979-01-01

    This picture shows one of the 2 new cavities installed in 1978-1979. The main RF-system of the SPS comprises four cavities: two of 20 m length and two of 16.5 m length. They are all installed in one long straight section (LSS 3). These cavities are of the travelling-wave type operating at a centre frequency of 200.2 MHz. They are wideband, filling time about 700 ns and untuned. The power amplifiers, using tetrodes are installed in a surface building 200 m from the cavities. Initially only two cavities were installed, a third cavity was installed in 1978 and a forth one in 1979. The number of power amplifiers was also increased: to the first 2 MW plant a second 2 MW plant was added and by end 1979 there were 8 500 kW units combined in pairs to feed each of the 4 cavities with up to about 1 MW RF power, resulting in a total accelerating voltage of about 8 MV. See also 7412016X, 7412017X, 7411048X

  9. Superconducting TESLA cavities

    Directory of Open Access Journals (Sweden)

    B. Aune

    2000-09-01

    Full Text Available The conceptional design of the proposed linear electron-positron collider TESLA is based on 9-cell 1.3 GHz superconducting niobium cavities with an accelerating gradient of E_{acc}≥25 MV/m at a quality factor Q_{0}≥5×10^{9}. The design goal for the cavities of the TESLA Test Facility (TTF linac was set to the more moderate value of E_{acc}≥15 MV/m. In a first series of 27 industrially produced TTF cavities the average gradient at Q_{0}=5×10^{9} was measured to be 20.1±6.2 MV/m, excluding a few cavities suffering from serious fabrication or material defects. In the second production of 24 TTF cavities, additional quality control measures were introduced, in particular, an eddy-current scan to eliminate niobium sheets with foreign material inclusions and stringent prescriptions for carrying out the electron-beam welds. The average gradient of these cavities at Q_{0}=5×10^{9} amounts to 25.0±3.2 MV/m with the exception of one cavity suffering from a weld defect. Hence only a moderate improvement in production and preparation techniques will be needed to meet the ambitious TESLA goal with an adequate safety margin. In this paper we present a detailed description of the design, fabrication, and preparation of the TESLA Test Facility cavities and their associated components and report on cavity performance in test cryostats and with electron beam in the TTF linac. The ongoing research and development towards higher gradients is briefly addressed.

  10. 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.)

  11. Cavity-enhanced spectroscopies

    CERN Document Server

    van Zee, Roger

    2003-01-01

    ""Cavity-Enhanced Spectroscopy"" discusses the use of optical resonators and lasers to make sensitive spectroscopic measurements. This volume is written by the researcchers who pioneered these methods. The book reviews both the theory and practice behind these spectroscopic tools and discusses the scientific discoveries uncovered by these techniques. It begins with a chapter on the use of optical resonators for frequency stabilization of lasers, which is followed by in-depth chapters discussing cavity ring-down spectroscopy, frequency-modulated, cavity-enhanced spectroscopy, intracavity spectr

  12. Tuned optical cavity magnetometer

    Science.gov (United States)

    Okandan, Murat; Schwindt, Peter

    2010-11-02

    An atomic magnetometer is disclosed which utilizes an optical cavity formed from a grating and a mirror, with a vapor cell containing an alkali metal vapor located inside the optical cavity. Lasers are used to magnetically polarize the alkali metal vapor and to probe the vapor and generate a diffracted laser beam which can be used to sense a magnetic field. Electrostatic actuators can be used in the magnetometer for positioning of the mirror, or for modulation thereof. Another optical cavity can also be formed from the mirror and a second grating for sensing, adjusting, or stabilizing the position of the mirror.

  13. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  14. 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....

  15. 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.

  16. 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.

  17. 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.

  18. Ambiguous walls

    DEFF Research Database (Denmark)

    Mody, Astrid

    2012-01-01

    The introduction of Light Emitting Diodes (LEDs) in the built environment has encouraged myriad applications, often embedded in surfaces as an integrated part of the architecture. Thus the wall as responsive luminous skin is becoming, if not common, at least familiar. Taking into account how wall...

  19. Double wall steam generator tubing

    International Nuclear Information System (INIS)

    Padden, T.R.; Uber, C.F.

    1983-01-01

    Double-walled steam generator tubing for the steam generators of a liquid metal cooled fast breeder reactor prevents sliding between the surfaces due to a mechanical interlock. Forces resulting from differential thermal expansion between the outer tube and the inner tube are insufficient in magnitude to cause shearing of base metal. The interlock is formed by jointly drawing the tubing, with the inside wall of the outer tube being already formed with grooves. The drawing causes the outer wall of the inner tube to form corrugations locking with the grooves. (author)

  20. 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

  1. Ambiguous walls

    DEFF Research Database (Denmark)

    Mody, Astrid

    2012-01-01

    The introduction of Light Emitting Diodes (LEDs) in the built environment has encouraged myriad applications, often embedded in surfaces as an integrated part of the architecture. Thus the wall as responsive luminous skin is becoming, if not common, at least familiar. Taking into account how walls...... have encouraged architectural thinking of enclosure, materiality, construction and inhabitation in architectural history, the paper’s aim is to define new directions for the integration of LEDs in walls, challenging the thinking of inhabitation and program. This paper introduces the notion...... of “ambiguous walls” as a more “critical” approach to design [1]. The concept of ambiguous walls refers to the diffuse status a lumious and possibly responsive wall will have. Instead of confining it can open up. Instead of having a static appearance, it becomes a context over time. Instead of being hard...

  2. SPS accelerating cavity

    CERN Multimedia

    CERN PhotoLab

    1976-01-01

    The SPS started up with 2 accelerating cavities (each consisting of 5 tank sections) in LSS3. They have a 200 MHz travelling wave structure (see 7411032 and 7802190) and 750 kW of power is fed to each of the cavities from a 1 MW tetrode power amplifier, located in a surface building above, via a coaxial transmission line. Clemens Zettler, builder of the SPS RF system, is standing at the side of one of the cavities. In 1978 and 1979 another 2 cavities were added and entered service in 1980. These were part of the intensity improvement programme and served well for the new role of the SPS as proton-antiproton collider. See also 7411032, 8011289, 8104138, 8302397.

  3. Dental Sealants Prevent Cavities

    Science.gov (United States)

    ... Digital Press Kit Read the MMWR Science Clips Dental Sealants Prevent Cavities Effective protection for children Language: ... more use of sealants and reimbursement of services. Dental care providers can Apply sealants to children at ...

  4. Statistical electromagnetics: Complex cavities

    NARCIS (Netherlands)

    Naus, H.W.L.

    2008-01-01

    A selection of the literature on the statistical description of electromagnetic fields and complex cavities is concisely reviewed. Some essential concepts, for example, the application of the central limit theorem and the maximum entropy principle, are scrutinized. Implicit assumptions, biased

  5. accelerating cavity from LEP

    CERN Multimedia

    This is an accelerating cavity from LEP, with a layer of niobium on the inside. Operating at 4.2 degrees above absolute zero, the niobium is superconducting and carries an accelerating field of 6 million volts per metre with negligible losses. Each cavity has a surface of 6 m2. The niobium layer is only 1.2 microns thick, ten times thinner than a hair. Such a large area had never been coated to such a high accuracy. A speck of dust could ruin the performance of the whole cavity so the work had to be done in an extremely clean environment. These challenging requirements pushed European industry to new achievements. 256 of these cavities are now used in LEP to double the energy of the particle beams.

  6. 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.

  7. Research on the HYLIFE liquid-first-wall concept for future laser-fusion reactors: liquid jet impact experiments. Final report No. 8

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1982-08-01

    The goal of this initial scoping study was to evaluate the transient and steady state drag of a single bar and of some selected arrays of bars and to determine the momentum removed from impacting liquid slugs. In order to achieve this aim, use has been made of both the published literature and experimental data obtained from a small-scale experimental apparatus. The implications of two possible scaling laws for use in designing the small-scale experiment are discussed. The use of near-universal curves to evaluate the momentum removed during the initial transient period is described. The small-scale apparatus used to obtain steady-state drag data is described. Finally, these results are applied to the HYLIFE fusion reactor

  8. 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

  9. 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.

  10. Hybrid vertical cavity laser

    DEFF Research Database (Denmark)

    Chung, Il-Sug; Mørk, Jesper

    2010-01-01

    A new hybrid vertical cavity laser structure for silicon photonics is suggested and numerically investigated. It incorporates a silicon subwavelength grating as a mirror and a lateral output coupler to a silicon ridge waveguide.......A new hybrid vertical cavity laser structure for silicon photonics is suggested and numerically investigated. It incorporates a silicon subwavelength grating as a mirror and a lateral output coupler to a silicon ridge waveguide....

  11. The Superconducting TESLA Cavities

    CERN Document Server

    Aune, B.; Bloess, D.; Bonin, B.; Bosotti, A.; Champion, M.; Crawford, C.; Deppe, G.; Dwersteg, B.; Edwards, D.A.; Edwards, H.T.; Ferrario, M.; Fouaidy, M.; Gall, P-D.; Gamp, A.; Gössel, A.; Graber, J.; Hubert, D.; Hüning, M.; Juillard, M.; Junquera, T.; Kaiser, H.; Kreps, G.; Kuchnir, M.; Lange, R.; Leenen, M.; Liepe, M.; Lilje, L.; Matheisen, A.; Möller, W-D.; Mosnier, A.; Padamsee, H.; Pagani, C.; Pekeler, M.; Peters, H-B.; Peters, O.; Proch, D.; Rehlich, K.; Reschke, D.; Safa, H.; Schilcher, T.; Schmüser, P.; Sekutowicz, J.; Simrock, S.; Singer, W.; Tigner, M.; Trines, D.; Twarowski, K.; Weichert, G.; Weisend, J.; Wojtkiewicz, J.; Wolff, S.; Zapfe, K.

    2000-01-01

    The conceptional design of the proposed linear electron-positron colliderTESLA is based on 9-cell 1.3 GHz superconducting niobium cavities with anaccelerating gradient of Eacc >= 25 MV/m at a quality factor Q0 > 5E+9. Thedesign goal for the cavities of the TESLA Test Facility (TTF) linac was set tothe more moderate value of Eacc >= 15 MV/m. In a first series of 27industrially produced TTF cavities the average gradient at Q0 = 5E+9 wasmeasured to be 20.1 +- 6.2 MV/m, excluding a few cavities suffering fromserious fabrication or material defects. In the second production of 24 TTFcavities additional quality control measures were introduced, in particular aneddy-current scan to eliminate niobium sheets with foreign material inclusionsand stringent prescriptions for carrying out the electron-beam welds. Theaverage gradient of these cavities at Q0 = 5E+9 amounts to 25.0 +- 3.2 MV/mwith the exception of one cavity suffering from a weld defect. Hence only amoderate improvement in production and preparation technique...

  12. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  13. 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.

  14. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  15. 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.

  16. Energy balance of algal biomass production in a 1-ha “Green Wall Panel” plant: How to produce algal biomass in a closed reactor achieving a high Net Energy Ratio

    International Nuclear Information System (INIS)

    Tredici, M.R.; Bassi, N.; Prussi, M.; Biondi, N.; Rodolfi, L.; Chini Zittelli, G.; Sampietro, G.

    2015-01-01

    Highlights: • Tetraselmis suecica production in a 1-ha GWP plant in Tuscany (Italy) has a NER < 1. • Major energy costs are embodied energy of GWP and mixing. • In a suitable location (North Africa) the NER increases by 40%. • Integration of photovoltaic in the GWP allows to achieve a NER of 1.7. • T. suecica cultivated in a GWP plant can yield up to 30 t of protein ha −1 year −1 . - Abstract: The annual productivity of Tetraselmis suecica in a 1-ha Green Wall Panel-II (GWP-II) plant in Tuscany (Italy) is 36 t (dry weight) ha −1 year −1 , which corresponds to an energy output of 799 GJ ha −1 year −1 . The energy inputs necessary to attain that productivity amount to 1362 GJ ha −1 year −1 , mainly given by the embodied energy of the reactor (about 30%), mixing (about 40%), fertilizers (11%) and harvesting (10%). The Net Energy Ratio (NER) of T. suecica production is thus 0.6. In a more suitable location (North Africa) productivity nearly doubles, reaching 66 t ha −1 year −1 , but the NER increases only by 40% and the gain (difference between output and inputs) remains negative. In a GWP-II integrated with photovoltaics (PV), the NER becomes 1.7 and the gain surpasses 600 GJ ha −1 year −1 . Marine microalgae cultivation in a GWP plant, in a suitable location, can attain high biomass productivities and protein yields 30 times higher than those achievable with traditional crops (soya). When the GWP reactor is integrated with PV, the process attains a positive energy balance, which substantially enhances its sustainability

  17. Wall Turbulence.

    Science.gov (United States)

    Hanratty, Thomas J.

    1980-01-01

    This paper gives an account of research on the structure of turbulence close to a solid boundary. Included is a method to study the flow close to the wall of a pipe without interferring with it. (Author/JN)

  18. 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)

  19. Construction of the LITL cavity structure

    International Nuclear Information System (INIS)

    Itoh, S.; Masuda, S.; Ukai, Y.; Hirao, Y.

    1984-01-01

    This report presents briefly the mechanical consideration for the 100 MHz four-vane RFQ (radio frequency quadrupole accelerator) structure construction. At first, the theoretical vane shape required to obtain the RFQ electric field distribution was determined. A numerically controlled milling machine was employed for the precise machining of the complicated shape. The data sets for NC machining and for checking the size of three-dimensional coordinates were made up. A small vane model was machined by way of trial experiment to check the data to verify the circular interpolation programmed NC machining method, and to investigate cutter interference. The errors in the measurement in machining were less than +- 30 micrometer. The resonator tank is 56 cm in inner diameter and 138 cm in length, and is made of mild steel of 35 mm thickness. The inside wall was plated with copper thickly. Various conditions for the copper plating were investigated. Four vanes were assembled within the cavity of the RFQ. The vanes were built in the cavity tank with high dimensional accuracy. It was a matter of primary concern to design acceptable mechanical rf joints and select suitable rf contact elements for a high Q value of the RFQ resonator cavity. Finally, the Q value was measured, and was 10,600. The cavity was able to be evacuated to 10 -7 Torr. (Kato, T.)

  20. 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

  1. X-ray imaging of superconducting radio frequency cavities

    Science.gov (United States)

    Musser, Susan Elizabeth

    The goal of this research was to develop an improved diagnostic technique to identify the location of defects that limit superconducting radio frequency (SRF) cavity performance during cavity testing or in existing accelerators. SRF cavities are primarily constructed of niobium. Electrons within the metal of a cavity under high electric field gradient have a probability of tunneling through the potential barrier. i e. leave the surface or are field emitted in regions where defects are encountered. Field emitted electrons are accelerated in the electric fields within the cavity. The electrons can have complicated trajectories and strike the cavity walls thus producing x-rays via Coulomb interactions and/or bremsstrahlung radiation. The endpoint energy of an x-ray spectrum predicts the electron maximum final kinetic energy within the cavity. Field emission simulations can then predict the source of the field-emitted electrons and the defect(s). In a multicell cavity the cells are coupled together and act as a set of coupled oscillators. There are multiple passbands of excitation for a multicell structure operating in a particular mode. For different passbands of operation the direction and amplitude of the fields within a cavity change from that of the normal accelerating mode. Field emitted electrons have different trajectories depending on the mode and thus produce x-rays in different locations. Using a collimated sodium iodide detector and subjecting a cavity to multiple passband modes at high electric field gradient the source of a cavity's x-rays can be determined. Knowing the location of the x-rays and the maximum electron kinetic energy; field emission simulations for different passband modes can be used to determine and verify the source of the field emitted electrons from mode to mode. Once identified, the defect(s) can be repaired or modifications made to the manufacturing process.

  2. 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.

  3. 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

  4. Video Toroid Cavity Imager

    Energy Technology Data Exchange (ETDEWEB)

    Gerald, Rex E. II; Sanchez, Jairo; Rathke, Jerome W.

    2004-08-10

    A video toroid cavity imager for in situ measurement of electrochemical properties of an electrolytic material sample includes a cylindrical toroid cavity resonator containing the sample and employs NMR and video imaging for providing high-resolution spectral and visual information of molecular characteristics of the sample on a real-time basis. A large magnetic field is applied to the sample under controlled temperature and pressure conditions to simultaneously provide NMR spectroscopy and video imaging capabilities for investigating electrochemical transformations of materials or the evolution of long-range molecular aggregation during cooling of hydrocarbon melts. The video toroid cavity imager includes a miniature commercial video camera with an adjustable lens, a modified compression coin cell imager with a fiat circular principal detector element, and a sample mounted on a transparent circular glass disk, and provides NMR information as well as a video image of a sample, such as a polymer film, with micrometer resolution.

  5. Metasurface external cavity laser

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Luyao, E-mail: luyaoxu.ee@ucla.edu; Curwen, Christopher A.; Williams, Benjamin S. [Department of Electrical Engineering, University of California, Los Angeles, California 90095 (United States); California NanoSystems Institute, University of California, Los Angeles, California 90095 (United States); Hon, Philip W. C.; Itoh, Tatsuo [Department of Electrical Engineering, University of California, Los Angeles, California 90095 (United States); Chen, Qi-Sheng [Northrop Grumman Aerospace Systems, Redondo Beach, California 90278 (United States)

    2015-11-30

    A vertical-external-cavity surface-emitting-laser is demonstrated in the terahertz range, which is based upon an amplifying metasurface reflector composed of a sub-wavelength array of antenna-coupled quantum-cascade sub-cavities. Lasing is possible when the metasurface reflector is placed into a low-loss external cavity such that the external cavity—not the sub-cavities—determines the beam properties. A near-Gaussian beam of 4.3° × 5.1° divergence is observed and an output power level >5 mW is achieved. The polarized response of the metasurface allows the use of a wire-grid polarizer as an output coupler that is continuously tunable.

  6. 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

  7. Earth-ionosphere cavity

    International Nuclear Information System (INIS)

    Tran, A.; Polk, C.

    1976-01-01

    To analyze ELF wave propagation in the earth-ionosphere cavity, a flat earth approximation may be derived from the exact equations, which are applicable to the spherical cavity, by introducing a second-order or Debye approximation for the spherical Hankel functions. In the frequency range 3 to 30 Hz, however, the assumed conditions for the Debye approximation are not satisfied. For this reason an exact evaluation of the spherical Hankel functions is used to study the effects of the flat earth approximation on various propagation and resonance parameters. By comparing the resonance equation for a spherical cavity with its flat earth counterpart and by assuming that the surface impedance Z/sub i/ at the upper cavity boundary is known, the relation between the eigenvalue ν and S/sub v/, the sine of the complex angle of incidence at the lower ionosphere boundary, is established as ν(ν + 1) = (kaS/sub v/) 2 . It is also shown that the approximation ν(ν + 1) approximately equals (ν + 1/2) 2 which was used by some authors is not adequate below 30 Hz. Numerical results for both spherical and planar stratification show that (1) planar stratification is adequate for the computation of the lowest three ELF resonance frequencies to within 0.1 Hz; (2) planar stratification will lead to errors in cavity Q and wave attenuation which increase with frequency; (3) computation of resonance frequencies to within 0.1 Hz requires the extension of the lower boundary of the ionosphere to a height where the ratio of conduction current to displacement current, (sigma/ωepsilon 0 ), is less than 0.3; (4) atmospheric conductivity should be considered down to ground level in computing cavity Q and wave attenuation

  8. 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

  9. Localization of a sound source in in a guided medium and reverberating field. Contribution to a study on leak localization in the internal wall of containment of a nuclear reactor in the case of a severe reactor accident

    International Nuclear Information System (INIS)

    Thomann, F.

    1996-01-01

    Basic data necessary for the localization of a leak in the internal wall of the containment are presented by studying the sound generated by gas jets coming out of (leaking fissures) as well as propagation in a guided medium. The results acquired have led us to choose the simple intercorrelation method and the matched filed processing method, both of which are likely to adequately handle our problems. Whereas the intercorrelation method appears to be limited in scope when dealing in the guided medium, the matched field processing is suited to leak localization over a surface of approximately 1000 m 2 (for a total surface of 10 000 m 2 ). Preliminary studies on the leak signal and on replicated signals have led us to limit the frequency band to 2600 - 3000 Hz. We have succeeded in locating a leak situated in an ordinary position with a minimum amount of replicated signals and basic data. We have improved on the estimation of Bartlett and MVDE (minimum variance distortion less filter) rendering them even more effective. Afterwards, we considered the severe accident situation and showed that the system can be installed in situ. (author)

  10. 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.

  11. Materials for superconducting cavities

    International Nuclear Information System (INIS)

    Bonin, B.

    1996-01-01

    The ideal material for superconducting cavities should exhibit a high critical temperature, a high critical field, and, above all, a low surface resistance. Unfortunately, these requirements can be conflicting and a compromise has to be found. To date, most superconducting cavities for accelerators are made of niobium. The reasons for this choice are discussed. Thin films of other materials such as NbN, Nb 3 Sn, or even YBCO compounds can also be envisaged and are presently investigated in various laboratories. It is shown that their success will depend critically on the crystalline perfection of these films. (author)

  12. Modeling of local steam condensation on walls in presence of non-condensable gases. Application to a loca calculation in reactor containment using the multidimensional geyser/tonus code

    Energy Technology Data Exchange (ETDEWEB)

    Benet, L.V.; Caroli, C.; Cornet, P. [Commissariat a l`Energie Atomique, Gif sur Yvette (France)] [and others

    1995-09-01

    This paper reports part of a study of possible severe pressurized water reactor (PWR) accidents. The need for containment modeling, and in particular for a hydrogen risk study, was reinforced in France after 1990, with the requirement that severe accidents must be taken into account in the design of future plants. This new need of assessing the transient local hydrogen concentration led to the development, in the Mechanical Engineering and Technology Department of the French Atomic Energy Commission (CEA/DMT), of the multidimensional code GEYSER/TONUS for containment analysis. A detailed example of the use of this code is presented. The mixture consisted of noncondensable gases (air or air plus hydrogen) and water vapor and liquid water. This is described by a compressible homogeneous two-phase flow model and wall condensation is based on the Chilton-Colburn formula and the analogy between heat and mass transfer. Results are given for a transient two-dimensional axially-symmetric computation for the first hour of a simplified accident sequence. In this there was an initial injection of a large amount of water vapor followed by a smaller amount and by hydrogen injection.

  13. Experimental investigation of cavity flows

    Energy Technology Data Exchange (ETDEWEB)

    Loeland, Tore

    1999-12-31

    This thesis uses LDV (Laser Doppler Velocimetry), PIV (Particle Image Velocimetry) and Laser Sheet flow Visualisation to study flow inside three different cavity configurations. For sloping cavities, the vortex structure inside the cavities is found to depend upon the flow direction past the cavity. The shape of the downstream corner is a key factor in destroying the boundary layer flow entering the cavity. The experimental results agree well with numerical simulations of the same geometrical configurations. The results of the investigations are used to find the influence of the cavity flow on the accuracy of the ultrasonic flowmeter. A method to compensate for the cavity velocities is suggested. It is found that the relative deviation caused by the cavity velocities depend linearly on the pipe flow. It appears that the flow inside the cavities should not be neglected as done in the draft for the ISO technical report on ultrasonic flowmeters. 58 refs., 147 figs., 2 tabs.

  14. Experimental investigation of cavity flows

    Energy Technology Data Exchange (ETDEWEB)

    Loeland, Tore

    1998-12-31

    This thesis uses LDV (Laser Doppler Velocimetry), PIV (Particle Image Velocimetry) and Laser Sheet flow Visualisation to study flow inside three different cavity configurations. For sloping cavities, the vortex structure inside the cavities is found to depend upon the flow direction past the cavity. The shape of the downstream corner is a key factor in destroying the boundary layer flow entering the cavity. The experimental results agree well with numerical simulations of the same geometrical configurations. The results of the investigations are used to find the influence of the cavity flow on the accuracy of the ultrasonic flowmeter. A method to compensate for the cavity velocities is suggested. It is found that the relative deviation caused by the cavity velocities depend linearly on the pipe flow. It appears that the flow inside the cavities should not be neglected as done in the draft for the ISO technical report on ultrasonic flowmeters. 58 refs., 147 figs., 2 tabs.

  15. 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

  16. 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.

  17. 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.

  18. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  19. Multipactors in klystron cavities

    International Nuclear Information System (INIS)

    Hayashi, Kazutaka; Iyeki, Hiroshi; Kikunaga, Toshiyuki.

    1993-01-01

    A multipactor phenomenon in a klystron causes gain shortage or instability problem. Some tests using a prototype klystron input cavity revealed the microwave discharges in vacuum with magnetic field. The test results and the methods to avoid multipactors are discussed in this paper. (author)

  20. What's a Cavity?

    Science.gov (United States)

    ... and deeper over time. Cavities are also called dental caries (say: KARE-eez), and if you have a ... made up mostly of the germs that cause tooth decay. The bacteria in your mouth make acids and when plaque clings to your teeth, the acids can eat away at the outermost ...

  1. Vertical cavity laser

    DEFF Research Database (Denmark)

    2016-01-01

    The present invention provides a vertical cavity laser comprising a grating layer comprising an in-plane grating, the grating layer having a first side and having a second side opposite the first side and comprising a contiguous core grating region having a grating structure, wherein an index...

  2. Oral cavity and jaw

    International Nuclear Information System (INIS)

    Solntsev, A.M.; Koval', G.Yu.

    1984-01-01

    Radioanatome of oral cavity and jaw is described. Diseases of the teeth, jaw, large salivary glands, temporo-mandibular articulation are considered. Roentgenograms of oral cacity and jaw of healthy people are presented and analyzed as well as roentgenograms in the above-mentioned diseases

  3. Niobium superconducting cavity

    CERN Multimedia

    CERN PhotoLab

    1980-01-01

    This 5-cell superconducting cavity, made from bulk-Nb, stems from the period of general studies, not all directed towards direct use at LEP. This one is dimensioned for 1.5 GHz, the frequency used at CEBAF and also studied at Saclay (LEP RF was 352.2 MHz). See also 7908227, 8007354, 8209255, 8210054, 8312339.

  4. Superconducting elliptical cavities

    CERN Document Server

    Sekutowicz, J K

    2011-01-01

    We give a brief overview of the history, state of the art, and future for elliptical superconducting cavities. Principles of the cell shape optimization, criteria for multi-cell structures design, HOM damping schemes and other features are discussed along with examples of superconducting structures for various applications.

  5. Additive Manufactured Superconducting Cavities

    Science.gov (United States)

    Holland, Eric; Rosen, Yaniv; Woolleet, Nathan; Materise, Nicholas; Voisin, Thomas; Wang, Morris; Mireles, Jorge; Carosi, Gianpaolo; Dubois, Jonathan

    Superconducting radio frequency cavities provide an ultra-low dissipative environment, which has enabled fundamental investigations in quantum mechanics, materials properties, and the search for new particles in and beyond the standard model. However, resonator designs are constrained by limitations in conventional machining techniques. For example, current through a seam is a limiting factor in performance for many waveguide cavities. Development of highly reproducible methods for metallic parts through additive manufacturing, referred to colloquially as 3D printing\\x9D, opens the possibility for novel cavity designs which cannot be implemented through conventional methods. We present preliminary investigations of superconducting cavities made through a selective laser melting process, which compacts a granular powder via a high-power laser according to a digitally defined geometry. Initial work suggests that assuming a loss model and numerically optimizing a geometry to minimize dissipation results in modest improvements in device performance. Furthermore, a subset of titanium alloys, particularly, a titanium, aluminum, vanadium alloy (Ti - 6Al - 4V) exhibits properties indicative of a high kinetic inductance material. This work is supported by LDRD 16-SI-004.

  6. Cavity Nesting Birds

    Science.gov (United States)

    Virgil E. Scott; Keith E. Evans; David R. Patton; Charles P. Stone

    1977-01-01

    Many species of cavity-nesting birds have declined because of habitat reduction. In the eastern United States, where primeval forests are gone, purple martins depend almost entirely on man-made nesting structures (Allen and Nice 1952). The hole-nesting population of peregrine falcons disappeared with the felling of the giant trees upon which they depended (Hickey and...

  7. LEP superconducting cavity

    CERN Multimedia

    1995-01-01

    Engineers work in a clean room on one of the superconducting cavities for the upgrade to the LEP accelerator, known as LEP-2. The use of superconductors allow higher electric fields to be produced so that higher beam energies can be reached.

  8. Open microwave cavities

    Czech Academy of Sciences Publication Activity Database

    Šeba, Petr; Rotter, I.; Mueller, M.; Persson, C.; Pichugin, Konstantin N.

    2001-01-01

    Roč. 9, - (2001), s. 484-487 ISSN 1386-9477 Institutional research plan: CEZ:A02/98:Z1-010-914 Keywords : microwave cavity * resonances Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.009, year: 2001

  9. Filling a Conical Cavity

    Science.gov (United States)

    Nye, Kyle; Eslam-Panah, Azar

    2016-11-01

    Root canal treatment involves the removal of infected tissue inside the tooth's canal system and filling the space with a dense sealing agent to prevent further infection. A good root canal treatment happens when the canals are filled homogeneously and tightly down to the root apex. Such a tooth is able to provide valuable service for an entire lifetime. However, there are some examples of poorly performed root canals where the anterior and posterior routes are not filled completely. Small packets of air can be trapped in narrow access cavities when restoring with resin composites. Such teeth can cause trouble even after many years and lead the conditions like acute bone infection or abscesses. In this study, the filling of dead-end conical cavities with various liquids is reported. The first case studies included conical cavity models with different angles and lengths to visualize the filling process. In this investigation, the rate and completeness at which a variety of liquids fill the cavity were observed to find ideal conditions for the process. Then, a 3D printed model of the scaled representation of a molar with prepared post spaces was used to simulate the root canal treatment. The results of this study can be used to gain a better understanding of the restoration for endodontically treated teeth.

  10. A Rare Tumor of Nasal Cavity: Glomangiopericytoma

    Directory of Open Access Journals (Sweden)

    Aysegul Verim

    2014-01-01

    Full Text Available Glomangiopericytoma is a rare vascular neoplasm characterized by a pattern of prominent perivascular growth. A 72-year-old woman was admitted to our clinic complaining of nasal obstruction, frequent epistaxis, and facial pain. A reddish tumor filling the left nasal cavity was observed on endoscopy and treated with endoscopic excision. Microscopically, closely packed cells interspersed with numerous thin-walled, branching staghorn vessels were seen. Glomangiopericytoma is categorized as a borderline low malignancy tumor by WHO classification. Long-term follow-up with systemic examination is necessary due to high risk of recurrence.

  11. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  12. Possible fusion reactor

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1976-05-01

    A scheme to improve performance characteristics of a tokamak-type fusion reactor is proposed. Basically, the tokamak-type plasma could be moved around so that the plasma could be heated by compression, brought to the region where the blanket surrounds the plasma, and moved so as to keep wall loading below the acceptable limit. This idea should be able to help to economize a fusion reactor

  13. Implosion of the small cavity and large cavity cannonball targets

    International Nuclear Information System (INIS)

    Nishihara, Katsunobu; Yamanaka, Chiyoe.

    1984-01-01

    Recent results of cannonball target implosion research are briefly reviewed with theoretical predictions for GEKKO XII experiments. The cannonball targets are classified into two types according to the cavity size ; small cavity and large cavity. The compression mechanisms of the two types are discussed. (author)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  15. 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.

  16. 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

  17. 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

  18. Surface processing for bulk niobium superconducting radio frequency cavities

    Science.gov (United States)

    Kelly, M. P.; Reid, T.

    2017-04-01

    The majority of niobium cavities for superconducting particle accelerators continue to be fabricated from thin-walled (2-4 mm) polycrystalline niobium sheet and, as a final step, require material removal from the radio frequency (RF) surface in order to achieve performance needed for use as practical accelerator devices. More recently bulk niobium in the form of, single- or large-grain slices cut from an ingot has become a viable alternative for some cavity types. In both cases the so-called damaged layer must be chemically etched or electrochemically polished away. The methods for doing this date back at least four decades, however, vigorous empirical studies on real cavities and more fundamental studies on niobium samples at laboratories worldwide have led to seemingly modest improvements that, when taken together, constitute a substantial advance in the reproducibility for surface processing techniques and overall cavity performance. This article reviews the development of niobium cavity surface processing, and summarizes results of recent studies. We place some emphasis on practical details for real cavity processing systems which are difficult to find in the literature but are, nonetheless, crucial for achieving the good and reproducible cavity performance. New approaches for bulk niobium surface treatment which aim to reduce cost or increase performance, including alternate chemical recipes, barrel polishing and ‘nitrogen doping’ of the RF surface, continue to be pursued and are closely linked to the requirements for surface processing.

  19. 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

  20. 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

  1. 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

  2. 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)

  3. Shielding wall for thermonuclear device

    International Nuclear Information System (INIS)

    Uchida, Takaho.

    1989-01-01

    This invention concerns shielding walls opposing to plasmas of a thermonuclear device and it is an object thereof to conduct reactor operation with no troubles even if a portion of shielding wall tiles should be damaged. That is, the shielding wall tiles are constituted as a dual layer structure in which the lower base tiles are connected by means of bolts to first walls. Further, the upper surface tiles are bolt-connected to the layer base tiles. In this structure, the plasma thermal loads are directly received by the surface layer tiles and heat is conducted by means of conduction and radiation to the underlying base tiles and the first walls. Even upon occurrence of destruction accidents to the surface layer tiles caused by incident heat or electromagnetic force upon elimination of plasmas, since the underlying base tiles remain as they are, the first walls constituted with stainless steels, etc. are not directly exposed to the plasmas. Accordingly, the integrity of the first walls having cooling channels can be maintained and sputtering intrusion of atoms of high atom number into the plasmas can be prevented. (I.S.)

  4. 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)

  5. Hollow waveguide cavity ringdown spectroscopy

    Science.gov (United States)

    Dreyer, Chris (Inventor); Mungas, Greg S. (Inventor)

    2012-01-01

    Laser light is confined in a hollow waveguide between two highly reflective mirrors. This waveguide cavity is used to conduct Cavity Ringdown Absorption Spectroscopy of loss mechanisms in the cavity including absorption or scattering by gases, liquid, solids, and/or optical elements.

  6. Optimization of photonic crystal cavities

    DEFF Research Database (Denmark)

    Wang, Fengwen; Sigmund, Ole

    2017-01-01

    We present optimization of photonic crystal cavities. The optimization problem is formulated to maximize the Purcell factor of a photonic crystal cavity. Both topology optimization and air-hole-based shape optimization are utilized for the design process. Numerical results demonstrate...... that the Purcell factor of the photonic crystal cavity can be significantly improved through optimization....

  7. 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

  8. 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.

  9. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2001-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  10. Colloquium: cavity optomechanics

    CERN Multimedia

    2011-01-01

    Monday 14 November 2011, 17:00 Ecole de Physique, Auditoire Stueckelberg Université de Genève Cavity optomechanics: controlling micro mechanical oscillators with laser light Prof. Tobias Kippenberg EPFL, Lausanne Laser light can be used to cool and to control trapped ions, atoms and molecules at the quantum level. This has lead to spectacular advances such as the most precise atomic clocks. An outstanding frontier is the control with lasers of nano- and micro-mechancial systems. Recent advances in cavity optomechanics have allowed such elementary control for the first time, enabling mechanical systems to be ground state cooled leading to readout with quantum limited sensitivity and permitting to explore new device concepts resulting from radiation pressure.  

  11. Three-dimensional self-consistent simulations of multipacting in superconducting radio frequency cavities. Final Report

    International Nuclear Information System (INIS)

    Nieter, Chet

    2010-01-01

    Superconducting radio frequency (SRF) cavities are a popular choice among researchers designing new accelerators because of the reduced power losses due to surface resistance. However, SRF cavities still have unresolved problems, including the loss of power to stray electrons. Sources of these electrons are field emission from the walls and ionization of background gas, but the predominant source is secondary emission yield (SEY) from electron impact. When the electron motion is in resonance with the cavity fields the electrons strike the cavity surface repeatedly creating a resonant build up of electrons referred to as multipacting. Cavity shaping has successfully reduced multipacting for cavities used in very high energy accelerators. However, multipacting is still a concern for the cavity power couplers, where shaping is not possible, and for cavities used to accelerate particles at moderate velocities. This Phase II project built upon existing models in the VORPAL simulation framework to allow for simulations of multipacting behavior in SRF cavities and their associated structures. The technical work involved allowed existing models of secondary electron generation to work with the complex boundary conditions needed to model the cavity structures. The types of data produced by VORPAL were also expanded to include data common used by cavity designers to evaluate cavity performance. Post-processing tools were also modified to provide information directly related to the conditions that produce multipacting. These new methods were demonstrated by running simulations of a cavity design being developed by researchers at Jefferson National Laboratory to attempt to identify the multipacting that would be an issue for the cavity design being considered. These simulations demonstrate that VORPAL now has the capabilities to assist researchers working with SRF cavities to understand and identify possible multipacting issues with their cavity designs.

  12. Leaching materials from cavities

    International Nuclear Information System (INIS)

    Hodgson, T.D.; Jordan, T.W.J.

    1980-01-01

    A material is leached from a cavity by contacting the material with a liquid and subjecting the liquid to a number of pressure cycles, each pressure cycle involving a decrease in pressure to cause boiling of the liquid, followed by a rise in pressure to inhibit the boiling. The method may include the step of heating the liquid to a temperature near to its boiling point. The material may be nuclear fuel pellets or calcium carbonate pellets. (author)

  13. Superconducting cavities for HERA

    International Nuclear Information System (INIS)

    Dwersteg, B.; Ebeling, W.; Moeller, W.D.; Renken, D.; Proch, D.; Sekutowicz, J.; Susta, J.; Tong, D.

    1988-01-01

    Superconducting 500 MHz cavities are developed to demonstrate the feasibility of upgrading the e-beam energy of the HERA storage ring. A prototype module with 2 x 4 cell resonators and appropriate fundamental and higher mode couplers has been designed at DESY and is being built by industrial firms. The design and results of RF and cryogenic measurements are reported in detail. 17 references, 10 figures, 2 tables

  14. 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

  15. 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

  16. 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)

  17. Nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

    1979-01-01

    Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

  18. 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.

  19. 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)

  20. 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