WorldWideScience

Sample records for large pool lmfbr

  1. Feasibility study on large pool-type LMFBR

    International Nuclear Information System (INIS)

    1984-01-01

    A feasibility study has been conducted from 1981 FY to 1983 FY, in order to evaluate the feasibility of a large pool-type LMFBR under the Japanese seismic design condition and safety design condition, etc. This study was aimed to establish an original reactor structure concept which meets those design conditions especially required in Japan. In the first year, preceding design concepts had been reviewed and several concepts were originated to be suitable to Japan. For typical two of them being selected by preliminary analysis, test programs were planned. In the second year, more than twenty tests with basic models had been conducted under severe conditions, concurrently analytical approaches were promoted. In the last year, larger model tests were conducted and analytical methods have been verified concerning hydrodynamic effects on structure vibration, thermo-hydraulic behaviours in reactor plena and so on. Finally the reactor structure concepts for a large pool-type LMFBR have been acknowledged to be feasible in Japan. (author)

  2. System seismic analysis of an innovative primary system for a large pool type LMFBR plant

    International Nuclear Information System (INIS)

    Pan, Y.C.; Wu, T.S.; Cha, B.K.; Burelbach, J.; Seidensticker, R.

    1984-01-01

    The system seismic analysis of an innovative primary system for a large pool type liquid metal fast breeder reactor (LMFBR) plant is presented. In this primary system, the reactor core is supported in a way which differs significantly from that used in previous designs. The analytical model developed for this study is a three-dimensional finite element model including one-half of the primary system cut along the plane of symmetry. The model includes the deck and deck mounted components,the reactor vessel, the core support structure, the core barrel, the radial neutron shield, the redan, and the conical support skirt. The sodium contained in the primary system is treated as a lumped mass appropriately distributed among various components. The significant seismic behavior as well as the advantages of this primary system design are discussed in detail

  3. A fundamental study on sodium-water reaction in the double-pool-type LMFBR

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Akimoto, Tokuzo

    1987-01-01

    In order to evaluate the pressure rise by large sodium-water reaction in the Double-Pool LMFBR, basic tests on pressure wave celerity in rectangular tube are carried out. The initial spike pressure in rectangular-shelled steam generator of the Double Pool reactor, strongly depends on pressure wave celerity. In this study, celerity was measured as a function of pressure wave rising time and pulse height, and influence of water around the test section on celerity was investigated. (author)

  4. Technical assessment study on pool-type LMFBR

    International Nuclear Information System (INIS)

    1986-01-01

    Technical assessment study on pool-type LMFBR was started in 1984 FY, inheriting the products from the Feasibility study, in order to accomplish cost reduction of reactor structure and enhanced structural reliability. This study consists of four major subjects; aseismic design development, component design optimization, high temperature structural design optimization and thermal hydraulics design optimization. In 1985 FY numbers of large model tests and analytical evaluations have been performed based on the prospects obtained in the first year's study. These tests and analyses have produced a lot of findings in each subject. They are concerning; (1) the effect of various building structures and analysis methods on floor response reduction, and data for evaluation of aseismic design concepts and structural integrity to seismic loading in the aseismic design development study. (2) data for evaluation of size reduction of main components in the reactor vessel, and heat transfer data required for structural integrity evaluation in the component design optimization study. (3) data for verification of inelastic analysis method, and assurance of technical applicability of disimilar weld in the high temperature structural design optimization study. (4) the effect of component size and location on thermal hydraulic characteristics, and data of thermal hydraulic similarity in thermal hydraulic design optimization study. This report summarizes the results obtained in 1985 FY. (author)

  5. Seismic isolation structure for pool-type LMFBR - isolation building with vertically isolated floor for NSSS

    International Nuclear Information System (INIS)

    Sakurai, A.; Shiojiri, H.; Aoyagi, S.; Matsuda, T.; Fujimoto, S.; Sasaki, Y.; Hirayama, H.

    1987-01-01

    The NSSS isolation floor vibration characteristics were made clear. Especially, the side support bearing (rubber bearing) is effective for horizontal floor motion restraint and rocking motion control. Seismic isolation effects for responses of the reactor components can be sufficiently expected, using the vertical seismic isolation floor. From the analytical and experimental studies, the following has been concluded: (1) Seismic isolation structure, which is suitable for large pool-type LMFBR, were proposed. (2) Seismic response characteristics of the seismic isolation structure were investigated. It was made clear that the proposed seismic isolation (Combination of the isolated building and the isolated NSSS floor) was effective. (orig./HP)

  6. An experimental study on sodium-water reaction in the double pool LMFBR, (4)

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo; Uotani, Masaki; Akimoto, Tokuzo

    1989-01-01

    Double Pool type LMFBR set the rectangular cross-sectional steam generator (SGs) inside a secondary vessel. The initial spike pressure rise caused by large sodium-water reaction in SGs might be radiated into a large sodium pool in the secondary vessel. Therefore basic experiments on pressure wave propagation were carried out by generating pressure wave in water by mean of a set of drop hummer and piston. But the experimental apparatus in water was not convenience to simulate the structure near the bottom end of the SGs shell. In this reports, experiments were carried out by generating pulse sound pressure in air, and compared with the results pressure waves in water. (author)

  7. Sodium-water reaction in double pool LMFBR, (5)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Kumagai, Hiromichi; Nishi, Yoshihisa; Uotani, Masaki

    1990-01-01

    Experiments were conducted using a 1/5 scale model of the Double Pool in order to evaluate a pressure rise caused by a large scale sodium-water reaction. The experiments were focused on the pressure rise caused by the piston motion of liquid sodium. It appeared from the results that the magnitude of this pressure rise depends on the depth of reaction point, and that a pressure rise more than 1 MPa would arise in the real Double Pool plant. A new design of steam generator is proposed to mitigate the pressure rise. (author)

  8. Key asset - inherent safety of LMFBR Pool Plant

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Lancet, R.T.; Mills, J.C.

    1984-04-01

    The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

  9. Monte-Carlo validation of secondary sodium activation in a pool-type LMFBR

    International Nuclear Information System (INIS)

    Plamiotti, G.; Rado, V.; Salvatores, M.

    1980-09-01

    The secondary sodium activation in a pool-type LMFBR is the main parameter to be accurately evaluated in the shield design. In the present work a complete two dimensional description of the system, including core, shielding and sodium up to Heat Exchangers, is coupled to local Heat Exchanger Monte-Carlo calculations. This refined calculation is used to deduce a simplified method to take into account the coupling of radial propagation in the Heat Exchanger and its finite cylindrical structure

  10. A fundamental study on sodium-water reaction in the double pool LMFBR, (3)

    International Nuclear Information System (INIS)

    Uotani, Masaki; Kumagai, Hiromichi; Nishi, Yoshihisa; Yoshida, Kazuo

    1989-01-01

    The double pool LMFBR is an innovative reactor that Central Research Institute of Electric Power Industry proposed for the purpose of reducing the construction cost of FBRs, and it is characterized by immersing steam generators in the annular plenum formed between the primary vessel and the outer secondary vessel. Therefore, it is expected that the pressure behavior at the time of sodium-water reaction due to the breaking of heating tubes is largely different from the case of steam generators of conventional FBRs. In order to ensure the soundness of the primary vessel that containes the reactor core, it is necessary to sufficiently grasp the pressure behavior in the plenum, and this basic experiment and analysis are related to the pressure behavior due to piston motion that arises in the initial period of quasi-steady pressure. About 1/10 scale annular plenum was used, and the generation of reaction product gas was simulated by the release of nitrogen. When gas was released in the plenum, the highest pressure rise occurred in the initial period of release, and thereafter, periodic variation arose. The pressure waveform and the value of pressure rise as the results of the model analysis agreed well with the measured results. (K.I.)

  11. Response matrix method for large LMFBR analysis

    International Nuclear Information System (INIS)

    King, M.J.

    1977-06-01

    The feasibility of using response matrix techniques for computational models of large LMFBRs is examined. Since finite-difference methods based on diffusion theory have generally found a place in fast-reactor codes, a brief review of their general matrix foundation is given first in order to contrast it to the general strategy of response matrix methods. Then, in order to present the general method of response matrix technique, two illustrative examples are given. Matrix algorithms arising in the application to large LMFBRs are discussed, and the potential of the response matrix method is explored for a variety of computational problems. Principal properties of the matrices involved are derived with a view to application of numerical methods of solution. The Jacobi iterative method as applied to the current-balance eigenvalue problem is discussed

  12. Study of structural attachments of a pool type LMFBR vessel through seismic analysis of a simplified three dimensional finite element model

    International Nuclear Information System (INIS)

    Ahmed, H.; Ma, D.

    1979-01-01

    A simplified three dimensional finite element model of a pool type LMFBR in conjunction with the computer program ANSYS is developed and scoping results of seismic analysis are produced. Through this study various structural attachments of a pool type LMFBR like the reactor vessel skirt support, the pump support and reactor shell-support structure interfaces are studied. This study also provides some useful results on equivalent viscous damping approach and some improvements to the treatment of equivalent viscous damping are recommended. This study also sets forth pertinent guidelines for detailed three dimensional finite element seismic analysis of pool type LMFBR

  13. Heat transfer performance of multilayer insulation system under roof slab of pool-type LMFBR

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Naohara, Nobuyuki; Uotani, Masaki

    1986-01-01

    To cope with thermal expansion of stainless steel plate, about 90 insulation structures are installed under the roof-slab of pool-type LMFBR. The objective of this study is to evaluate from heat transfer experiment and visualized experiment, the effect of distance between each thermal insulation structure on heat transfer characteristics of insulation system under roof-slab. Two types of insulation structures are selected, one is open type and the other is closed type. Distance between each thermal insulation structure and hot surface temperatures are varied as a parameter. Furthermore, heat flux of the roof-slab insulation system of reactor are estimated from the results of heat transfer experiment. (author)

  14. Stability of inner baffle-shell of pool type LMFBR - experimental and theoretical studies

    International Nuclear Information System (INIS)

    Lebey, J.; Combescure, A.

    1987-01-01

    I pool type LMFBR, the primary coolant circuit, inside the main vessel, comprises a hot plenum separated from a cold plenum by an inner baffle. For Superphenix 1 reactor, it was judged advisable to built a double-shell baffle, each shell withstanding only one type of loading (primary loading for one shell, secondary loading for the other). Due to the size and intricacy of the structure, this design involves unnegligible supplementary costs and manufacturing difficulties. Thus, an alternative solution has been studied for future plants projects. It consists of a single shell baffle having a shape especially studied to sustain the two types of applied loadings (thermal plus primary loadings). Such a shape was calculated by NOVATOME, and it was decided to check the ability of methods of analysis to predict the ruin of this structure under primary loading. For this purpose, a mock-up has been tested, and the experimental results compared with the calculated ones. (orig./GL)

  15. An internal core catcher for a pool L.M.F.B.R. and connected studies

    International Nuclear Information System (INIS)

    Le Rigoleur, C.; Kayser, G.

    1979-01-01

    This paper describes an internal core catcher for a pool LMFBR. Problems related to retention of debris are studied: downward progression of debris from the core to the core catcher, debris bed formation, heat transfer below the core catcher plate and to the main vessel, mechanical resistance. These results are used to estimate the performances of the internal core catcher for a given core melt-down-accident. It is seen that for a uniform thickness layer on the core catcher the retention capabilities are satisfactory. Then the problem of a heap of debris is approached. Dryout is studied. Uncertainties related to the bed characteristics and problems of extended dryout beds are put forward

  16. Large sodium pool fires in a closed or in a naturally vented cell

    International Nuclear Information System (INIS)

    Rzekiecki, R.; Malet, J.C.; Sophy, Y.; Joly, C.; Claverie, J.

    1986-01-01

    Within the framework of R and D studies related to LMFBR sodium handling safety, a facility named ESMERALDA 1 was provided. This permits full scale demonstration of the control of sodium fires. ESMERALDA is a French-Italian project including EDF, CEA, NERSA, NOVATOME, partners. Large pool fires studies are a part of it. (author)

  17. Model of large pool fires

    Energy Technology Data Exchange (ETDEWEB)

    Fay, J.A. [Department of Mechanical Engineering, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)]. E-mail: jfay@mit.edu

    2006-08-21

    A two zone entrainment model of pool fires is proposed to depict the fluid flow and flame properties of the fire. Consisting of combustion and plume zones, it provides a consistent scheme for developing non-dimensional scaling parameters for correlating and extrapolating pool fire visible flame length, flame tilt, surface emissive power, and fuel evaporation rate. The model is extended to include grey gas thermal radiation from soot particles in the flame zone, accounting for emission and absorption in both optically thin and thick regions. A model of convective heat transfer from the combustion zone to the liquid fuel pool, and from a water substrate to cryogenic fuel pools spreading on water, provides evaporation rates for both adiabatic and non-adiabatic fires. The model is tested against field measurements of large scale pool fires, principally of LNG, and is generally in agreement with experimental values of all variables.

  18. Model of large pool fires

    International Nuclear Information System (INIS)

    Fay, J.A.

    2006-01-01

    A two zone entrainment model of pool fires is proposed to depict the fluid flow and flame properties of the fire. Consisting of combustion and plume zones, it provides a consistent scheme for developing non-dimensional scaling parameters for correlating and extrapolating pool fire visible flame length, flame tilt, surface emissive power, and fuel evaporation rate. The model is extended to include grey gas thermal radiation from soot particles in the flame zone, accounting for emission and absorption in both optically thin and thick regions. A model of convective heat transfer from the combustion zone to the liquid fuel pool, and from a water substrate to cryogenic fuel pools spreading on water, provides evaporation rates for both adiabatic and non-adiabatic fires. The model is tested against field measurements of large scale pool fires, principally of LNG, and is generally in agreement with experimental values of all variables

  19. Seismic analysis of large pools

    Energy Technology Data Exchange (ETDEWEB)

    Dong, R.G.; Tokarz, F.J.

    1976-11-17

    Large pools for storing spent, nuclear fuel elements are being proposed to augment present storage capacity. To preserve the ability to isolate portions of these pools, a modularization requirement appears desirable. The purpose of this project was to investigate the effects of modularization on earthquake resistance and to assess the adequacy of current design methods for seismic loads. After determining probable representative pool geometries, three rectangular pool configurations, all 240 x 16 ft and 40 ft deep, were examined. One was unmodularized; two were modularized into 80 x 40 ft cells in one case and 80 x 80 ft cells in the other. Both embedded and above-ground installations for a hard site and embedded installations for an intermediate hard site were studied. It was found that modularization was unfavorable in terms of reducing the total structural load attributable to dynamic effects, principally because one or more cells could be left unfilled. The walls of unfilled cells would be subjected to significantly higher loads than the walls of a filled, unmodularized pool. Generally, embedded installations were preferable to above-ground installations, and the hard site was superior to the intermediate hard site. It was determined that Housner's theory was adequate for calculating hydrodynamic effects on spent fuel storage pools. Current design methods for seismic loads were found to be satisfactory when results from these methods were compared with those from LUSH analyses. As a design method for dynamic soil pressure, we found the Mononobe-Okabe theory, coupled with correction factors as suggested by Seed, to be acceptable. The factors we recommend for spent fuel storage pools are tabulated.

  20. Seismic analysis of large pools

    International Nuclear Information System (INIS)

    Dong, R.G.; Tokarz, F.J.

    1976-01-01

    Large pools for storing spent, nuclear fuel elements are being proposed to augment present storage capacity. To preserve the ability to isolate portions of these pools, a modularization requirement appears desirable. The purpose of this project was to investigate the effects of modularization on earthquake resistance and to assess the adequacy of current design methods for seismic loads. After determining probable representative pool geometries, three rectangular pool configurations, all 240 x 16 ft and 40 ft deep, were examined. One was unmodularized; two were modularized into 80 x 40 ft cells in one case and 80 x 80 ft cells in the other. Both embedded and above-ground installations for a hard site and embedded installations for an intermediate hard site were studied. It was found that modularization was unfavorable in terms of reducing the total structural load attributable to dynamic effects, principally because one or more cells could be left unfilled. The walls of unfilled cells would be subjected to significantly higher loads than the walls of a filled, unmodularized pool. Generally, embedded installations were preferable to above-ground installations, and the hard site was superior to the intermediate hard site. It was determined that Housner's theory was adequate for calculating hydrodynamic effects on spent fuel storage pools. Current design methods for seismic loads were found to be satisfactory when results from these methods were compared with those from LUSH analyses. As a design method for dynamic soil pressure, we found the Mononobe-Okabe theory, coupled with correction factors as suggested by Seed, to be acceptable. The factors we recommend for spent fuel storage pools are tabulated

  1. Research program on the feasibility of pool type LMFBR in Japan

    International Nuclear Information System (INIS)

    Hattori, Sadao

    1982-01-01

    The Central Research Institute of Electric Power Industry has started the feasibility study to evaluate the possiblity of existence of large pool type FBR plants in Japan as the three-year project from fiscal 1981. The development of FBRs is indispensable for the effective use of nuclear fuel and the establishment of energy security. The knowledge on the characteristics of FBR core, sodium technology and others has advanced rapidly in Japan. At the stage of the practical reactors with large capacity, the pool type is naturally considered as the object of selection, but the aseismatic capability and safety of the large containment vessels for the pool type and the qualitative and quantitative acceptability of the research and development for the pool type are the problems. The difference between the loop type and the pool type is only the structural change arising from the difference in the arrangement of equipment. The pool type reactors have been operated already in the UK and France. The objective of the research and main subjects, the total plan and research organization, the fundamental condition of investigation, the research procedure for respective subjects, and the outline of model test are discribed. The change of design and safety standards in the future must be predicted and taken in consideration in the research. (Kako, I.)

  2. Study on the seismic response of reactor vessel of pool type LMFBR including fluid-structure interaction

    International Nuclear Information System (INIS)

    Tanimoto, K.; Ito, T.; Fujita, K.; Kurihara, C.; Sawada, Y.; Sakurai, A.

    1988-01-01

    The paper presents the seismic response of reactor vessel of pool type LMFBR with fluid-structure interaction. The reactor vessel has bottom support arrangement, the same core support system as Super-Phenix in France. Due to the bottom support arrangement, the level of core support is lower than that of the side support arrangement. So, in this reactor vessel, the displacement of the core top tends to increase because of the core's rocking. In this study, we investigated the vibration and seismic response characteristics of the reactor vessel. Therefore, the seismic experiments were carried out using one-eighth scale model and the seismic response including FSI and sloshing were investigated. From this study, the effect of liquid on the vibration characteristics and the seismic response characteristics of reactor vessel were clarified and sloshing characteristics were also clarified. It was confirmed that FEM analysis with FSI can reproduce the seismic behavior of the reactor vessel and is applicable to seismic design of the pool type LMFBR with bottom support arrangement. (author). 5 refs, 14 figs, 2 tabs

  3. Large-scale pool fires

    Directory of Open Access Journals (Sweden)

    Steinhaus Thomas

    2007-01-01

    Full Text Available A review of research into the burning behavior of large pool fires and fuel spill fires is presented. The features which distinguish such fires from smaller pool fires are mainly associated with the fire dynamics at low source Froude numbers and the radiative interaction with the fire source. In hydrocarbon fires, higher soot levels at increased diameters result in radiation blockage effects around the perimeter of large fire plumes; this yields lower emissive powers and a drastic reduction in the radiative loss fraction; whilst there are simplifying factors with these phenomena, arising from the fact that soot yield can saturate, there are other complications deriving from the intermittency of the behavior, with luminous regions of efficient combustion appearing randomly in the outer surface of the fire according the turbulent fluctuations in the fire plume. Knowledge of the fluid flow instabilities, which lead to the formation of large eddies, is also key to understanding the behavior of large-scale fires. Here modeling tools can be effectively exploited in order to investigate the fluid flow phenomena, including RANS- and LES-based computational fluid dynamics codes. The latter are well-suited to representation of the turbulent motions, but a number of challenges remain with their practical application. Massively-parallel computational resources are likely to be necessary in order to be able to adequately address the complex coupled phenomena to the level of detail that is necessary.

  4. Seismic analysis of a large LMFBR with fluid-structure interactions

    International Nuclear Information System (INIS)

    Ma, D.C.

    1985-01-01

    The seismic analysis of a large LMFBR with many internal components and structures is presented. Both vertical and horizontal seismic excitations are considered. The important hydrodynamic phenomena such as fluid-structure interaction, sloshing, fluid coupling and fluid inertia effects are included in the analysis. The results of this study are discussed in detail. Information which is useful to the design of future reactions under seismic conditions is also given. 4 refs., 12 figs

  5. Development of concept and neutronic calculation method for large LMFBR core

    International Nuclear Information System (INIS)

    Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.

    1991-01-01

    Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)

  6. Analytical treatment of large leak pressure behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Hori, Masao; Miyake, Osamu

    1980-07-01

    Simplified analytical methods applicable to the estimation of initial pressure spike in case of a large leak accident in LMFBR steam generators were devised as follows; (i) Estimation of the initial water leak rate by the centered rarefaction wave method, (ii) Estimation of the initial pressure spike by the one-dimensional compressible method with either the columnar bubble growth model or the spherical bubble growth model. These methods were compared with relevant experimental data or other more elaborate analyses and validated to be usable in simple geometry and limited time span cases. Application of these methods to an actual steam generator case was explained and demonstrated. (author)

  7. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

    2003-01-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

  8. Large sodium water reaction calculations in a LMFBR steam generator

    International Nuclear Information System (INIS)

    Finck, P.; Lepareux, M.; Schwab, B.; Blanchet, Y.

    1986-05-01

    The French approach to the analysis of large and violent sodium water reactions is presented. The basis for choosing the Design Basis Accident is discussed. An energetical analysis of the physical phenomena involved stresses the specific needs for computing tools. The feature of these tools are then described, and a validation test is presented. Finally, industrial applications are described. 8 refs

  9. Heat transfer performance of multi-layer insulation structure under roof-slab of pool-type LMFBR

    International Nuclear Information System (INIS)

    Kinoshita, I.; Yoshida, K.; Uotani, M.; Fukada, T.

    1988-01-01

    At the normal operation of the pool-type LMFBR, the free surface of liquid sodium at about 500 0 C is present below the roof-slab, separated by a space of the argon cover gas. The temperature of the roof-slab has to be maintained low and uniform in the horizontal direction for sufficient strength of the structure. Therefore, thermal insulation structures must be installed on the lower surface of the roof-slab. In addition to the installation of thermal insulator, forced cooling of the roof-slab is required for assured structural integrity of the roof-slab. The capacity of cooling equipment can be reduced by installation of structures with high thermal insulating performance. The objective of this study is to evaluate the thermal insulation characteristics of multi-layer type insulator installed below the roof-slab by analytically and experimentally. The analytical study is intended to evaluate the effect of number, distance and emissivity of layers on the heat transfer performances. This is treated as the one-dimensional heat transfer with natural convection, conduction and thermal radiation. In the experiments, we have evaluated effects of gap distances between adjacent thermal insulators placed below the roof-slab on the thermal insulation performances

  10. Manipulator for fuel assemblies in a spent fuel pool, especially for a LMFBR

    International Nuclear Information System (INIS)

    Dalmas, R.

    1988-01-01

    The spent fuel manipulator has - a travelling crane moving longitudinally: - a carriage moving on the travelling crane in a direction perpendicular to its motion so that the carriage is positioned over each assembly, - a telescopic rod carried by the carriage and terminating in a vertically mobile grapple, - a tubular shielded hood on the carriage extending downwards to house the rod, grapple and fuel assembly and maintaining a biologically acceptable level of radiation above the surface of the pool [fr

  11. Influence of reactor design on the establishment of natural circulation in pool-type LMFBR

    International Nuclear Information System (INIS)

    Durham, M.E.

    1976-01-01

    The general principles involved in establishing natural circulation in a pool-type liquid metal cooled fast breeder reactor following loss of a.c. supplies are elucidated and the effects of design features by use of the computer code MELANI are quantified. It is shown that natural circulation can provide a feasible means of emergency core cooling in addition to that provided by pony motors. The choice of primary pump rundown time has a significant effect in controlling peak core outlet temperatures in the hypothetical case of natural circulation alone being the core heat removal process. (author)

  12. Large-scale tests of aqueous scrubber systems for LMFBR vented containment

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.

    1980-01-01

    Six large-scale air cleaning tests performed in the Containment Systems Test Facility (CSTF) are described. The test conditions simulated those postulated for hypothetical accidents in an LMFBR involving containment venting to control hydrogen concentration and containment overpressure. Sodium aerosols were generated by continously spraying sodium into air and adding steam and/or carbon dioxide to create the desired Na 2 O 2 , Na 2 CO 3 or NaOH aerosol. Two air cleaning systems were tested: (a) spray quench chamber, educator venturi scrubber and high efficiency fibrous scrubber in series; and (b) the same except with the spray quench chamber eliminated. The gas flow rates ranged up to 0.8 m 3 /s (1700 acfm) at temperatures to 313 0 C (600 0 F). Quantities of aerosol removed from the gas stream ranged up to 700 kg per test. The systems performed very satisfactorily with overall aerosol mass removal efficiencies exceeding 99.9% in each test

  13. New approach to the design of core support structures for large LMFBR plants

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions

  14. Measurements in large JP-4 pool fires

    International Nuclear Information System (INIS)

    Keltner, N.R.; Kent, L.A.; Schneider, M.E.

    1987-01-01

    Over the past four years, Sandia National Laboratories has conducted a number of large pool fire tests to evaluate the design of radioactive material (RAM) shipping containers. Some of these tests have been designed to define the thermal environment and some have been used for certification testing. In each test there have been a number of fire diagnostic measurements. The simplest sets of diagnostics have involved measurements of temperature at several elevations on arrays of towers, measurements of hot wall heat flux with small calorimeters suspended from the towers, the average fuel recession rate, and the wind speed and direction. The most complex sets of diagnostics have included the above and in various tests added radiometers in the lower flame zone, centerline velocity measurements at a number of elevations, radiometers and calorimeters at the fuel surface, large cylindrical and flat plate calorimeters, infrared imaging, time resolved fuel recession rates, and a variety of soot particle concentration and size measurements made in the plume with a tethered balloon and an instrumented airplane. All of the large fires have been conducted in a 9.1 m by 18.3 m pool using JP-4 as the fuel. Typical duration is one-half hour. Covering all of the results is beyond the scope of a single paper. Conditionally sampled temperature and velocity measurements from one fire will be presented; for this fire, a 20 cm layer of fuel was floated on 61 cm of water. Pool surface heat flux, fuel recession rate data, and smoke emission data from a second fire are given. Because the wind has a strong effect on the temperature and velocity measurements, conditional sampling has been used to try to obtain data during periods of low winds. 10 refs., 3 figs

  15. Component design for LMFBR's

    International Nuclear Information System (INIS)

    Fillnow, R.H.; France, L.L.; Zerinvary, M.C.; Fox, R.O.

    1975-01-01

    Just as FFTF has prototype components to confirm their design, FFTF is serving as a prototype for the design of the commercial LMFBR's. Design and manufacture of critical components for the FFTF system have been accomplished primarily using vendors with little or no previous experience in supplying components for high temperature sodium systems. The exposure of these suppliers, and through them a multitude of subcontractors, to the requirements of this program has been a necessary and significant step in preparing American industry for the task of supplying the large mechanical components required for commercial LMFBR's

  16. Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

    1982-01-01

    Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

  17. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  18. Seismic sloshing experiments of large pool-type fast breeder reactors

    International Nuclear Information System (INIS)

    Sakurai, A.; Masuko, Y.; Kurihara, C.; Ishihama, K.; Yashiro, T.; Rodwell, E.

    1989-01-01

    This paper presents the results of seismic sloshing experiments performed on large pool-type LMFBR vessels. Two types of tests were performed. The first type of test was designed to understand the basis phenomena of sloshing (limited to linear sloshing only) and evaluate the effects of the deck-mounted components (i.e., IHXs, pumps, and UIS) on sloshing wave heights using a 1/10-scale model (diameter 2.23 m x H 1.03 m) of the LSPB 1340 MWe pool plant. The second type of test was designed to evaluate the structural integrity of the thermal baffles of the roof-deck to withstand sloshing impulsive pressures (focused on nonlinear sloshing), using a two-dimensional 1/3-scale model (L 8 m x W 3 m x H 2.6 m) of a typical 1000 MWe pool plant. The results of the linear sloshing tests have shown that: 1. the vessel wall stiffness has no effect on the sloshing natural frequency; 2. sloshing wave heights are lowered by 30% to 50% in the presence of the deck-mounted components; and 3. damping factors of sloshing are not influenced by the wall stiffness while they are increased by the presence of the deck-mounted components. The results of the nonlinear sloshing tests are that: 1. the maximum impulsive pressure occurs when the first effective wave strikes at the roof-deck, and thereafter the impulsive pressure decreases irrespective of the impact velocity of the fluid; 2. the first effective wave refers to the case in which the height of the fluid free surface becomes nearly twice the height of the cover gas space; and 3. the structural integrity of the thermal baffles for the roof-deck against the sloshing load was confirmed. In addition to these results, two sloshing-caused problems were identified. The first one is the spillover of hot sodium into the gas-dam type thermal insulator. The second one is cover-gas entrainment into sodium which might lead to a transient overpower (TOP) incident because of the presence of gas bubbles in the reactor core. (orig./HP)

  19. Review of events at large pool-type irradiators

    International Nuclear Information System (INIS)

    Trager, E.A. Jr.

    1989-03-01

    Large pool-type gamma irradiators are used in applications such as the ''cold'' sterilization of medical and pharmaceutical supplies, and recent changes in federal regulations make it possible they will be used extensively in the preservation of foodstuffs. Because of this possible large increase in the use of irradiators, the Office of Nuclear Materials Safety and Safeguards was interested in knowing what events had occurred at irradiators. The event data would be used as background in developing new regulations on irradiators. Therefore, AEOD began a study of the operating experience at large, wet source storage gamma irradiators. The scope of the study was to assess all available operating information on large (≥ 250,000 curie), pool-type irradiators licensed by both the NRC and the Agreement States, and events at foreign facilities. The study found that about 0.12 events have been reported per irradiator-year. Most of these events were precursor events, in that there was no evidence of damage to the radioactive sources or degradation in the level of safety of the facility. Events with more significant impacts had a reported frequency of about 0.01 event per irradiator-year. However, the actual rate of occurrence of events of concern to the staff may be higher because there are few specific reporting requirements for events at irradiators. It is suggested that during development of a regulation for large pool-type irradiators consideration be given to specifying requirements for: reporting breakdowns in access control systems; periodic inspection of the source movement and suspension system; systems to detect source leakage and product contamination; allowable pool leakage; and feedback of information on operational events involving safety-important systems

  20. Hardware concepts for a large low-energetics LMFBR core. Final report

    International Nuclear Information System (INIS)

    Hutter, E.; Batch, R.V.

    1980-12-01

    A design study was made to identify a practical set of hardware configurations that would embody the requirements developed in the numerical study of a low-energetics core and blanket for a prototype large breeder reactor. Dimensioned drawings are presented for fuel, blanket, reflector/shield, and control rod subassemblies. A horizontal cross section drawing shows how these subassemblies are arranged in the total core/blanket assembly. A core support is illustrated showing a dual plenums arrangement

  1. Pressure transients resulting from sodium-water reaction following a large leak in LMFBR steam generator

    International Nuclear Information System (INIS)

    Rajput, A.K.

    1984-01-01

    The study of sodium water reaction, following a large leak, concerns primarily with the estimation of pressure/flow transients that are developed in the steam generator and the associated secondary circuit. This paper describes the mathematical formulations used in SWRT (Sodium Water Reaction Transients) code developed to estimate such pressure transients for FBTR plant. The results, obtained using SWRT have been presented for a leak in economiser (20m from bottom water header) and for a leak in super heater portions. A time lag of 50 m sec was considered for rupture disc takes to burst once the pressure experienced by it exceeds the set value. Also described in annexure to this paper is the mathematical formulation for two phase transient flow for the better estimation of leak rate from the ruptured end of the damaged heat transfer tube. This leak model considers slip but assumes thermal equilibrium between the liquid and vapour phases

  2. Measurements in large pool fires with an actively cooled calorimeter

    International Nuclear Information System (INIS)

    Koski, J.A.; Wix, S.D.

    1995-01-01

    The pool fire thermal test described in Safety Series 6 published by the International Atomic Energy Agency (IAEA) or Title 10, Code of Federal Regulations, Part 71 (10CFR71) in the United States is one of the most difficult tests that a container for larger ''Type B'' quantities of nuclear materials must pass. If retests of a container are required, costly redesign and project delays can result. Accurate measurements and modeling of the pool fire environment will ultimately lower container costs by assuring that containers past the pool fire test on the first attempt. Experiments indicate that the object size or surface temperature of the container can play a role in determining local heat fluxes that are beyond the effects predicted from the simple radiative heat transfer laws. An analytical model described by Nicolette and Larson 1990 can be used to understand many of these effects. In this model a gray gas represents soot particles present in the flame structure. Close to the container surface, these soot particles are convectively and radiatively cooled and interact with incident energy from the surrounding fire. This cooler soot cloud effectively prevents some thermal radiation from reaching the container surface, reducing the surface heat flux below the value predicted by a transparent medium model. With some empirical constants, the model suggested by Nicolette and Larson can be used to more accurately simulate the pool fire environment. Properly formulated, the gray gas approaches also fast enough to be used with standard commercial computer codes to analyze shipping containers. To calibrate this type of model, accurate experimental measurements of radiative absorption coefficients, flame temperatures, and other parameters are necessary. A goal of the calorimeter measurements described here is to obtain such parameters so that a fast, useful design tool for large pool fires can be constructed

  3. Workshop on large molten pool heat transfer summary and conclusions

    International Nuclear Information System (INIS)

    1994-01-01

    The CSNI Workshop on Large Molten Heat Transfer held at Grenoble (France) in March 1994 was organised by CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases (PWG4) with the cooperation of the Principal Working Group on Coolant System Behaviour (FWG2) and in collaboration with the Grenoble Nuclear Research Centre of the French Commissariat a l'Energie Atomique (CEA). Conclusions and recommendations are given for each of the five sessions of the workshops: Feasibility of in-vessel core debris cooling through external cooling of the vessel; Experiments on molten pool heat transfer; Calculational efforts on molten pool convection; Heat transfer to the surrounding water - experimental techniques; Future experiments and ex-vessel studies (open forum discussion)

  4. The Phoenix series large scale LNG pool fire experiments.

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, Richard B.; Jensen, Richard Pearson; Demosthenous, Byron; Luketa, Anay Josephine; Ricks, Allen Joseph; Hightower, Marion Michael; Blanchat, Thomas K.; Helmick, Paul H.; Tieszen, Sheldon Robert; Deola, Regina Anne; Mercier, Jeffrey Alan; Suo-Anttila, Jill Marie; Miller, Timothy J.

    2010-12-01

    The increasing demand for natural gas could increase the number and frequency of Liquefied Natural Gas (LNG) tanker deliveries to ports across the United States. Because of the increasing number of shipments and the number of possible new facilities, concerns about the potential safety of the public and property from an accidental, and even more importantly intentional spills, have increased. While improvements have been made over the past decade in assessing hazards from LNG spills, the existing experimental data is much smaller in size and scale than many postulated large accidental and intentional spills. Since the physics and hazards from a fire change with fire size, there are concerns about the adequacy of current hazard prediction techniques for large LNG spills and fires. To address these concerns, Congress funded the Department of Energy (DOE) in 2008 to conduct a series of laboratory and large-scale LNG pool fire experiments at Sandia National Laboratories (Sandia) in Albuquerque, New Mexico. This report presents the test data and results of both sets of fire experiments. A series of five reduced-scale (gas burner) tests (yielding 27 sets of data) were conducted in 2007 and 2008 at Sandia's Thermal Test Complex (TTC) to assess flame height to fire diameter ratios as a function of nondimensional heat release rates for extrapolation to large-scale LNG fires. The large-scale LNG pool fire experiments were conducted in a 120 m diameter pond specially designed and constructed in Sandia's Area III large-scale test complex. Two fire tests of LNG spills of 21 and 81 m in diameter were conducted in 2009 to improve the understanding of flame height, smoke production, and burn rate and therefore the physics and hazards of large LNG spills and fires.

  5. Modelling of decay heat removal using large water pools

    International Nuclear Information System (INIS)

    Munther, R.; Raussi, P.; Kalli, H.

    1992-01-01

    The main task for investigating of passive safety systems typical for ALWRs (Advanced Light Water Reactors) has been reviewing decay heat removal systems. The reference system for calculations has been represented in Hitachi's SBWR-concept. The calculations for energy transfer to the suppression pool were made using two different fluid mechanics codes, namely FIDAP and PHOENICS. FIDAP is based on finite element methodology and PHOENICS uses finite differences. The reason choosing these codes has been to compare their modelling and calculating abilities. The thermal stratification behaviour and the natural circulation was modelled with several turbulent flow models. Also, energy transport to the suppression pool was calculated for laminar flow conditions. These calculations required a large amount of computer resources and so the CRAY-supercomputer of the state computing centre was used. The results of the calculations indicated that the capabilities of these codes for modelling the turbulent flow regime are limited. Output from these codes should be considered carefully, and whenever possible, experimentally determined parameters should be used as input to enhance the code reliability. (orig.). (31 refs., 21 figs., 3 tabs.)

  6. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1979-03-01

    This document contains up-to-date data on existing or firmly decided prototype or demonstration LMFBR reactors (Table I), on planned commercial size LMFBR according to the present status of design (Table II) and on experimental fast reactors such as BOR-60, DFR, EBR-II, FERMI, FFTF, JOYO, KNK-II, PEC, RAPSODIE-FORTISSIMO (Table III). Only corrected and revised parameters submitted by the countries participating in the IWGFR are included in this document

  7. Pool fires in a large scale ventilation system

    International Nuclear Information System (INIS)

    Smith, P.R.; Leslie, I.H.; Gregory, W.S.; White, B.

    1991-01-01

    A series of pool fire experiments was carried out in the Large Scale Flow Facility of the Mechanical Engineering Department at New Mexico State University. The various experiments burned alcohol, hydraulic cutting oil, kerosene, and a mixture of kerosene and tributylphosphate. Gas temperature and wall temperature measurements as a function of time were made throughout the 23.3m 3 burn compartment and the ducts of the ventilation system. The mass of the smoke particulate deposited upon the ventilation system 0.61m x 0.61m high efficiency particulate air filter for the hydraulic oil, kerosene, and kerosene-tributylphosphate mixture fires was measured using an in situ null balance. Significant increases in filter resistance were observed for all three fuels for burning time periods ranging from 10 to 30 minutes. This was found to be highly dependent upon initial ventilation system flow rate, fuel type, and flow configuration. The experimental results were compared to simulated results predicted by the Los Alamos National Laboratory FIRAC computer code. In general, the experimental and the computer results were in reasonable agreement, despite the fact that the fire compartment for the experiments was an insulated steel tank with 0.32 cm walls, while the compartment model FIRIN of FIRAC assumes 0.31 m thick concrete walls. This difference in configuration apparently caused FIRAC to consistently underpredict the measured temperatures in the fire compartment. The predicted deposition of soot proved to be insensitive to ventilation system flow rate, but the measured values showed flow rate dependence. However, predicted soot deposition was of the same order of magnitude as measured soot deposition

  8. CFD modeling of pool swell during large break LOCA

    International Nuclear Information System (INIS)

    Yan, Jin; Bolger, Francis; Li, Guangjun; Mintz, Saul; Pappone, Daniel

    2009-01-01

    GE had conducted a series of one-third scale three-vent air tests in support the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface has been tracked by conductivity probes. There are many pressure monitors inside the test rig. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient 3-Dimensional CFD model of the one-third scale Mark III suppression pool swell process is constructed. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to those from the test. Through the comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked. (author)

  9. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  10. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  11. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  12. DATA-POOL : a direct-access data base for large-scale nuclear codes

    International Nuclear Information System (INIS)

    Yamano, Naoki; Koyama, Kinji; Naito, Yoshitaka; Minami, Kazuyoshi.

    1991-12-01

    A direct-access data base DATA-POOL has been developed for large-scale nuclear codes. The data can be stored and retrieved with specifications of simple node names, by using the DATA-POOL access package written in the FORTRAN 77 language. A management utility POOL for the DATA-POOL is also provided. A typical application of the DATA-POOL is shown to the RADHEAT-V4 code system developed for performing safety analyses of radiation shielding. Many samples and error messages are also noted to apply the DATA-POOL for the other code systems. This report is provided for a manual of DATA-POOL. (author)

  13. Role of fuel bubble phenomenology in assessment of LMFBR source term

    International Nuclear Information System (INIS)

    Cho, D.H.; Condiff, D.W.; Chan, S.H.

    1985-01-01

    Phenomenological aspects of a fuel vapor bubble formed in the sodium pool in a hypothetical severe accident are considered. The potential for fuel bubble collapse in the sodium pool is analyzed. It appears that for a wide range of hypothetical LMFBR accidents involving core vaporization, the fuel vapor bubble would likely be quenched and collapse prior to migration to the cover gas region. Such rapid quenching is due mainly to radiative heat transfer from the fuel bubble, coupled with the inherent capability of the sodium pool (large subcooling and high thermal conductivity) to dissipate thermal energy. Major uncertainty in the analysis concerns fuel vapor condensation phenomena at the sodium interface and its effect on the sodium surface radiation absorptivity. This is discussed in detail

  14. Seismic loads in modularized and unmodularized large pools located on hard or intermediate hard sites

    Energy Technology Data Exchange (ETDEWEB)

    Dong, R G [California Univ., Livermore (USA). Lawrence Livermore Lab.

    1977-12-01

    To augment the present capacity of pools for storing spent nuclear fuel elements, pools larger than those in current use are being planned. These pools may or may not be modularized into cells. Because of the large size of the pools, seismic loads are of significant interest. In particular, the effects of modularization and site hardness are of concern. The study presented in this paper reveals that modularization is generally unfavourable, because it creates the option of leaving one or more cells empty which in turn results in higher structural loads. The wall which separates a filled cell from an empty cell, or the wall which bears against earth on one side and faces an empty cell on the other, becomes very highly stressed. For the particular pool geometries examined, a hard site is generally preferred over an intermediate hard site in terms of structural loads.

  15. LMFBR: safety aspects

    International Nuclear Information System (INIS)

    Natta, M.

    1990-01-01

    This presentation of LMFBR safety is limited at Super Phenix reactor. After a brief description of the reactor, some details on safety systems, in normal or accidental conditions, are given. The main functions studied are: chain reaction trip, residual power evacuation, reactor containment. In heavy accident the behaviour of Super Phenix is studied which its particular characteristics and the possibilities of operators reactions. The probability of appearance and the maximum consequences of heavy accidents are given [fr

  16. Status of the LMFBR development

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, J J

    1975-01-01

    The development of any new power generation system which can make a major contribution to our energy needs is a multi-faceted task involving the utilization of major human and material resources. The LMFBR development, which has the potential for supplying abundant energy for generations, is therefore a large, multi-faceted program. This summary will cover (1) the need for the liquid metal fast breeder reactor, (2) an overall perspective of its development throughout the world, (3) a brief look at the in-depth technological development program in the United States, (4) a description and status of the two major projects now under way in the program, the Fast Flux Test Facility and the Clinch River Breeder Reactor Plant, and (5) a review of the plans for continued development to achieve a reliable, safe and economic power generation system for practical commercial use on the utility networks of the country.

  17. Formation and maintenance of a forced pool-riffle couplet following loading of large wood

    Science.gov (United States)

    Thompson, D. M.; Fixler, S. A.

    2017-11-01

    Pool-riffle maintenance has been documented in numerous studies, but it has been almost impossible to characterize detailed natural pool-riffle formation mechanisms because of the lack of baseline data prior to pool establishment. In 2013, a study was conducted on the Blackledge River in Connecticut to document the formation of a new pool-riffle couplet on a section of river that had previously been studied from 1999 to 2001. In 2001, the study reach contained a scour hole with a residual depth of 0.08 ± 0.09 m downstream of a 1930s paired deflector with no identifiable riffle immediately downstream. At this time, a large, severely undercut, hemlock tree was noted along the left bank. Sometime between fall 2001 and 2004, the tree fell perpendicular to flow across the channel and formed a large wood (LW) jam and new pool-riffle couplet several meters downstream of the old scour hole. Pool spacing along the reach decreased from 4.47 bankfull widths (BFW) in 1999 to 3.83 BFW after the new pool-riffle couplet formed. The new pool has a residual depth, the water depth of the streambed depression below the elevation of the immediate downstream hydraulic control, of 1.36 ± 0.075 to 1.59 ± 0.075 m, which resulted from a combination of 1.32 ± 0.09 m or less of incision below the old scour hole (95.6% or less of the depth increase) and up to 0.18 ± 0.09 m of downstream deposition and associated backwater formation (13.2% or less of the depth increase). To assess dynamic stability of the pool-riffle couplet over several flood cycles, surficial fine-sediment and organic material along the reach were quantified. The 23-m-long pool stores 25.7% of the surficial fine grained sediments and 15.4% of organic material along a 214-m-long reach that includes one additional artificially created pool. An adjacent 50-m-long secondary channel impacted by the LW jam stores 65.3% of the surficial fine-grained sediments and 54.8% of organic material along the full reach.

  18. A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

    International Nuclear Information System (INIS)

    Hartung, J.; Berk, S.

    1978-01-01

    Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

  19. LMFBR plant parameters

    International Nuclear Information System (INIS)

    1985-07-01

    This document has been prepared on the basis of information compiled by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains parameters of 25 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBR). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA) are presented because its design was nearly finished and most of the components were fabricated at the time when the project was terminated. Three reactors (RAPSODIE (France), DFR (UK) and EFFBR (USA)) have been shut down. However, they are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. The first LMFBRs (CLEMENTINE (USA), EBR-1 (USA), BR-2 (USSR), BR-5 (USSR)) and very special reactors (LAMPRE (USA), SEFOR (USA)) were not recommended by the members of the IWGFR to be included in the report

  20. Analysis of sodium pool fire in SFEF for assessing the limiting pool fire

    International Nuclear Information System (INIS)

    Mangarjuna Rao, P.; Ramesh, S.S.; Nashine, B.K.; Kasinathan, N.; Chellapandi, P.

    2011-01-01

    Accidental sodium leaks and resultant sodium fires in Liquid Metal Fast Breeder Reactor (LMFBR) systems can create a threat to the safe operation of the plant. To avoid this defence-in depth approach is implemented from the design stage of reactor itself. Rapid detection of sodium leak and fast dumping of the sodium into the storage tank of a defective circuit, leak collection trays, adequate lining of load bearing structural concrete and extinguishment of the sodium fire are the important defensive measures in the design, construction and operation of a LMFBR for protection against sodium leaks and their resultant fires. Evaluation of sodium leak events and their consequences by conducting large scale engineering experiments is very essential for effective implementation of the above protection measures for sodium fire safety. For this purpose a Sodium Fire Experimental Facility (SFEF) is constructed at SED, IGCAR. SFEF is having an experimental hall of size 9 m x 6 m x 10 m with 540 m 3 volume and its design pressure is 50 kPa. It is a concrete structure and provided with SS 304 liner, which is fixed to the inside surfaces of walls, ceiling and floor. A leak tight door of size (1.8 m x 2.0 m) is provided to the experimental hall and the facility is provided with a sodium equipment hall and a control room. Experimental evaluation of sodium pool fire consequences is an important activity in the LMFBR sodium fire safety related studies. An experimental program has been planned for different types of sodium fire studies in SFEF. A prior to that numerical analysis have been carried out for enclosed sodium pool fires using SOFIRE-II sodium pool fire code for SFEF experimental hall configuration to evaluate the limiting pool fire. This paper brings out results of the analysis carried out for this purpose. Limiting pool fire of SFEF depends on the exposed surface area of the pool, amount of sodium in the pool, oxygen concentration and initial sodium temperature. Limiting

  1. Impact of LMFBR operating experience on PFBR design

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chetal, S.C.; Chellapandi, P.; Govindarajan, S.; Lee, S.M.; Kameswara Rao, A.S.L.; Prabhakar, R.; Raghupathy, S.; Sodhi, B.S.; Sundaramoorthy, T.R.; Vaidyanathan, G.

    2000-01-01

    PFBR is a 500 MWe, sodium cooled, pool type, fast breeder reactor currently under detailed design. It is essential to reduce the capital cost of PFBR in order to make it competitive with thermal reactors. Operating experience of LMFBRs provides a vital input towards simplification of the design, improving its reliability, enhancing safety and achieving overall cost reduction. This paper includes a summary of LMFBR operating experience and details the design features of PFBR as influenced by operating experience of LMFBRs. (author)

  2. Shielding design method for LMFBR validation on the Phenix factor

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  3. Voluntary rewards mediate the evolution of pool punishment for maintaining public goods in large populations

    Science.gov (United States)

    Sasaki, Tatsuya; Uchida, Satoshi; Chen, Xiaojie

    2015-03-01

    Punishment is a popular tool when governing commons in situations where free riders would otherwise take over. It is well known that sanctioning systems, such as the police and courts, are costly and thus can suffer from those who free ride on other's efforts to maintain the sanctioning systems (second-order free riders). Previous game-theory studies showed that if populations are very large, pool punishment rarely emerges in public good games, even when participation is optional, because of second-order free riders. Here we show that a matching fund for rewarding cooperation leads to the emergence of pool punishment, despite the presence of second-order free riders. We demonstrate that reward funds can pave the way for a transition from a population of free riders to a population of pool punishers. A key factor in promoting the transition is also to reward those who contribute to pool punishment, yet not abstaining from participation. Reward funds eventually vanish in raising pool punishment, which is sustainable by punishing the second-order free riders. This suggests that considering the interdependence of reward and punishment may help to better understand the origins and transitions of social norms and institutions.

  4. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  5. Safety consequences of local initiating events in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

  6. Safety consequences of local initiating events in an LMFBR

    International Nuclear Information System (INIS)

    Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.; Wang, P.Y.

    1975-12-01

    The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered

  7. Technical considerations relative to removal of sodium from LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, J S; Asquith, J G

    1975-07-01

    Reviewed in this paper are technical considerations which are of importance in choosing between an alcohol process and a moist nitrogen process for the removal of sodium from LMFBR components. Results observed in laboratory tests and in the cleaning of large scale components (e.g. a 28 MWt Modular Steam Generator Test Unit) are presented and discussed. (author)

  8. Cold pool organization and the merging of convective updrafts in a Large Eddy Simulation

    Science.gov (United States)

    Glenn, I. B.; Krueger, S. K.

    2016-12-01

    Cold pool organization is a process that accelerates the transition from shallow to deep cumulus convection, and leads to higher deep convective cloud top heights. The mechanism by which cold pool organization enhances convection remains not well understood, but the basic idea is that since precipitation evaporation and a low equivalent potential temperature in the mid-troposphere lead to strong cold pools, the net cold pool effect can be accounted for in a cumulus parameterization as a relationship involving those factors. Understanding the actual physical mechanism at work will help quantify the strength of the relationship between cold pools and enhanced deep convection. One proposed mechanism of enhancement is that cold pool organization leads to reduced distances between updrafts, creating a local environment more conducive to convection as updrafts entrain parcels of air recently detrained by their neighbors. We take this hypothesis one step further and propose that convective updrafts actually merge, not just exchange recently processed air. Because entrainment and detrainment around an updraft draws nearby air in or pushes it out, respectively, they act like dynamic flow sources and sinks, drawing each other in or pushing each other away. The acceleration is proportional to the inverse square of the distance between two updrafts, so a small reduction in distance can make a big difference in the rate of merging. We have shown in previous research how merging can be seen as collisions between different updraft air parcels using Lagrangian Parcel Trajectories (LPTs) released in a Large Eddy Simulation (LES) during a period with organized deep convection. Now we use a Eulerian frame of reference to examine the updraft merging process during the transition from shallow to organized deep convection. We use a case based on the Large-Scale Biosphere-Atmosphere Experiment in Amazonia (LBA) for our LES. We directly measure the rate of entrainment and the properties

  9. LMFBR plant parameters 1991

    International Nuclear Information System (INIS)

    1991-03-01

    The document has been prepared on the basis of information provided by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains updated parameters of 27 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBRs). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA), PEC (Italy), RAPSODIE (France), DFR (UK) and EFFBR (USA) are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. Two more reactors appeared in the list: European Fast Reactor (EFR) and PRISM (USA). Parameters of these reactors included in this publication are based on the data from the papers presented at the 23rd Annual Meeting of the IWGFR. All in all more than four hundred corrections and additions have been made to update the document. The report is intended for specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors

  10. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  11. Measurement of temperature distributions in large pool fires with the use of directional flame thermometers

    International Nuclear Information System (INIS)

    Koski, Jorman A.

    2000-01-01

    Temperatures inside the flame zone of large regulatory pool fires measured during tests of radioactive materials packages vary widely with both time and position. Measurements made with several Directional Flame Thermometers, in which a thermocouple is attached to a thin metal sheet that quickly approaches flame temperatures, have been used to construct fire temperature distributions and cumulative probability distributions. As an aid to computer simulations of these large fires, these distributions are presented. The distributions are constructed by sorting fire temperature data into bins 10 C wide. A typical fire temperature distribution curve has a gradual increase starting at about 600 C, with the number of observations increasing to a peak near 1000 C, followed by an abrupt decrease in frequency, with no temperatures observed above 1200 C

  12. Effects of thin-layer boilover on flame geometry and dynamics in large hydrocarbon pool fires

    Energy Technology Data Exchange (ETDEWEB)

    Ferrero, Fabio; Munoz, Miguel; Arnaldos, Josep [Centre d' Estudis del Risc Tecnologic (CERTEC), Chemical Engineering Department, Universitat Politecnica de Catalunya, Diagonal 647, 08028-Barcelona, Catalonia (Spain)

    2007-03-15

    This work aims to estimate the effects of thin-layer boilover on flame geometry and dynamics. A series of large scale experiments (in pools ranging from 1.5 to 6 m in diameter) were performed using gasoline and diesel as fuel. As expected, only diesel showed evidence of this phenomenon. This article presents a summary of the results obtained for flame height, tilt and pulsation. Flame height increases during water ebullition, though the increase is no longer detectable when wind speed exceeds certain values. Correlations previously presented in the literature to predict flame length and tilt were modified in order to fit the results obtained during thin-layer boilover. However, the influence on flame tilt is not as great and the equations for the stationary period seem suitable for the entire fire. Results of flame pulsation during the stationary period fill the gap in the literature for fires between 1.5 and 6 m and fit previous correlations. On the other hand, during ebullition, the flame pulsates faster, as air entrainment is greater and, as one would expect, this effect decreases with pool size. A new equation for estimating pulsation frequency during boilover is proposed. (author)

  13. Pooling knowledge and improving safety for contracted works at a large industrial park.

    Science.gov (United States)

    Agnello, Patrizia; Ansaldi, Silvia; Bragatto, Paolo

    2015-01-01

    At a large chemical park maintenance is contracted by the major companies operating the plants to many small firms. The cultural and psychological isolation of contractor workers was recognized a root cause of severe accidents in the recent years. That problem is common in chemical industry. The knowledge sharing has been assumed a good key to involve contractors and sub contractors in safety culture and contributing to injuries prevention. The selection of personal protective equipment PPE for the maintenance works has been taken as benchmark to demonstrate the adequateness of the proposed approach. To support plant operators, contractors and subcontractors in PPE discussion, a method has been developed. Its core is a knowledge-base, organized in an Ontology, as suitable for inferring decisions. By means of this tool all stakeholders have merged experience and information and find out the right PPE, to be provided, with adequate training and information package. PPE selection requires sound competencies about process and environmental hazards, including major accident, preventive and protective measures, maintenance activities. These pieces of knowledge previously fragmented among plant operators and contractors, have to be pooled, and used to find out the adequate PPE for a number of maintenance works. The PPE selection is per se important, but it is also a good chance to break the contractors' isolation and involve them in safety objectives. Thus by pooling experience and practical knowledge, the common understanding of safety issues has been strengthened.

  14. Web service discovery among large service pools utilising semantic similarity and clustering

    Science.gov (United States)

    Chen, Fuzan; Li, Minqiang; Wu, Harris; Xie, Lingli

    2017-03-01

    With the rapid development of electronic business, Web services have attracted much attention in recent years. Enterprises can combine individual Web services to provide new value-added services. An emerging challenge is the timely discovery of close matches to service requests among large service pools. In this study, we first define a new semantic similarity measure combining functional similarity and process similarity. We then present a service discovery mechanism that utilises the new semantic similarity measure for service matching. All the published Web services are pre-grouped into functional clusters prior to the matching process. For a user's service request, the discovery mechanism first identifies matching services clusters and then identifies the best matching Web services within these matching clusters. Experimental results show that the proposed semantic discovery mechanism performs better than a conventional lexical similarity-based mechanism.

  15. Sodium mists behavior in cover gas space of an LMFBR

    International Nuclear Information System (INIS)

    Himeno, Y.; Takahashi, J.

    1978-03-01

    This paper present the sodium mist behavior in Argon cover gas space of an LMFBR experimentally using a test vessel of 1,400 mm in axial length, 305.5 mm in inner diameter and about 100 l in volume. Experiments are consisted with measurements of the mist concentration and the mist gravitational settling flux between the sodium pool temperature range of 290 0 to 520 0 C. The results are discussed under the monosize assumption of the particles, and the particle sizes and evaporation rate are derived. Transient and steady state mist concentration behavior were also investigated. (author)

  16. Simulation of LMFBR pump transients and comparison to LOF that occurred at EBR-II

    International Nuclear Information System (INIS)

    Koenig, F.F.; Dean, E.M.

    1985-01-01

    In a large LMFBR plant design, a number of pumps in parallel will feed the core. It must be demonstrated that the plant can continue to operate with the loss of one of the primary pumps. It is desirable not to have check valves in the loop from a reliability and economic standpoint. Simulations have been made to determine the consequences of a loss of one pump in a four-loop pool plant in which no plant protection action is taken. This analysis would be used to determine the required power rundown that would accompany pump loss. The two primary centrifugal pumps in EBR-II feed the core and blanket plenums in two parallel flow paths. The loss of one pump will result in decrease core flow and reverse flow through the down pump since no check valves are present in the system. For a large pool plant with four primary pumps, the loss of one pump will also result in reverse flow through the down pump if check valves of flow diodes are not included. The resulting flow transient has been modeled for EBR-II and the large plant using the DNSP program

  17. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  18. Future development LMFBR-steam generators SNR2

    International Nuclear Information System (INIS)

    Essebaggers, J.; Pors, J.G.

    1975-01-01

    The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

  19. LMFBR Blanket Physics Project progress report No. 2

    International Nuclear Information System (INIS)

    Forbes, I.A.; Driscoll, M.J.; Rasmussen, N.C.; Lanning, D.D.; Kaplan, I.

    1971-01-01

    This is the second annual report of an experimental program for the investigation of the neutronics of benchmark mock-ups of LMFBR blankets. Work was devoted primarily to measurements on Blanket Mock-Up No. 2, a simulation of a typical large LMFBR radial blanket and its steel reflector. Activation traverses and neutron spectra were measured in the blanket; calculations of activities and spectra were made for comparison with the measured data. The heterogeneous self-shielding effect for 238 U capture was found to be the most important factor affecting the comparison. Optimization and economic studies were made which indicate that the use of a high-albedo reflector material such as BeO or graphite may improve blanket neutronics and economics

  20. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  1. An analytical framework for common-pool resource–large technical system (CPR-LTS constellations

    Directory of Open Access Journals (Sweden)

    Pär Blomkvist

    2013-02-01

    Full Text Available This paper introduces an analytical framework for a special phenomenon: when a common-pool resource (CPR institution and a large technical system (LTS are connected and mutually interdependent. The CPR in this case is a node managed by its appropriators within a centrally planned and managed system; here named CPR-LTS constellations. Our framework is empirically derived from two historical investigations of CPR institutions within two LTSs, the agricultural-technical system and the road transport system of Sweden. By comparing similarities and differences it is possible to identify paths to successes and failures. To understand why one survived and the other disappeared we connect Elinor Ostrom’s theories about management of CPRs with Thomas P. Hughes’s theories about LTSs. We are proposing a framework that can bridge the gap between theories about management of CPRs and LTSs. By combining the two theories it should be possible to better understand how small-scale producers using bottom-up CPRs can be linked to top-down LTSs.We will argue that to fit within an LTS, a CPR needs alignment between different parts or components within the constellation/system and alignment with other systems and institutions in society. We propose three analytical levels to deal with the phenomenon of aligning a CPR project to an existing, large sociotechnical system:Local alignment (CPR: How are CPRs organized and managed at local sites?Sociotechnical alignment (CPR-LTS: How are CPRs connected to the sociotechnical system?Contextual alignment: How are CPR-LTS constellations aligned with neighboring institutions and systems in society?Our work indicates that for successful management of a CPR-LTS constellation it is important that the CPR be included in legislation and that government agencies support the CPR in alignment with the LTS. Legislators must recognize the CPR-part in the CPR-LTS constellation so that its institutional body is firmly established in

  2. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  3. Expansion joints for LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Dzenus, M.; Hundhausen, W.; Jansing, W.

    1979-10-15

    This discourse recounts efforts put into the SNR-2 project; specifically the development of compensation devices. The various prototypes of these compensation devices are described and the state of development reviewed. The expansion joints were developed on the basis of specific design criteria whereby differentiation is made between expansion joints of small and large nominal diameter. Expansion joints for installation in the sodium-filled primary piping are equipped with safety bellows in addition to the actual working bellows.

  4. LMFBR safety criteria: cost-benefit considerations under the constraint of an a priori risk criterion

    International Nuclear Information System (INIS)

    Hartung, J.

    1979-01-01

    The role of cost-benefit considerations and a priori risk criteria as determinants of Core Disruptive Accident (CDA)-related safety criteria for large LMFBR's is explored with the aid of quantitative risk and probabilistic analysis methods. A methodology is described which allows a large number of design and siting alternatives to be traded off against each other with the goal of minimizing energy generation costs subject to the constraint of both an a priori risk criterion and a cost-benefit criterion. Application of this methodology to a specific LMFBR design project is described and the results are discussed. 5 refs

  5. Use of reliability in the LMFBR industry

    International Nuclear Information System (INIS)

    Penland, J.R.; Smith, A.M.; Goeser, D.K.

    1977-01-01

    This mission of a Reliability Program for an LMFBR should be to enhance the design and operational characteristics relative to safety and to plant availability. Successful accomplishment of this mission requires proper integration of several reliability engineering tasks--analysis, testing, parts controls and program controls. Such integration requires, in turn, that the program be structured, planned and managed. This paper describes the technical integration necessary and the management activities required to achieve mission success for LMFBR's

  6. Structural and containment response to LMFBR accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

    1978-01-01

    The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

  7. Comparison of different LMFBR primary containment codes applied to a Benchmark problem

    International Nuclear Information System (INIS)

    Benuzzi, A.

    1986-01-01

    The Cont Benchmark calculation exercise is a project sponsored by the Containment Loading and Response Group, a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee - CEC. A full-size typical Pool type LMFBR undergoing a postulated Core Disruptive Accident (CDA) has been defined by Belgonucleaire-Brussels under a study contract financed by the CEC and has been submitted to seven containment code calculations. The results of these calculations are presented and discussed in this paper

  8. LMFBR system-wide transient analysis: the state of the art and US validation needs

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

    1982-01-01

    This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants

  9. Expansion joints for LMFBR

    International Nuclear Information System (INIS)

    Dzenus, M.; Hundhausen, W.; Jansing, W.

    1980-01-01

    This discourse recounts efforts put into the SNR-2 project; specifically the development of compensation devices. The various prototypes of these compensation devices are described and the state of the development reviewed. Large Na (sodium)-heat transfer systems require a lot of valuable space if the component lay-out does not include compensation devices. So, in order to condense the spatial requirement as much as possible, expansion joints must be integrated into the pipe system. There are two basic types to suit the purpose: axial expansion joints and angular expansion joints. The expansion joints were developed on the basis of specific design criteria whereby differentiation is made between expansion joints of small and large nominal diameter. Expansion joints for installation in the sodium-filled primary piping are equipped with safety bellows in addition to the actual working bellows. Expansion joints must be designed and mounted in a manner to completely withstand seismic forces. The design must exclude any damage to the bellows during intermittent operations, that is, when sodium is drained the bellows' folds must be completely empty; otherwise residual solidified sodium could destroy the bellows when restarting. The expansion joints must be engineered on the basis of the following design data for the secondary system of the SNR project: working pressure: 16 bar; failure mode pressure: 5 events; failure mode: 5 sec., 28.5 bar, 520 deg. C; working temperature: 520 deg. C; temperature transients: 30 deg. C/sec.; service life: 200,000 h; number of load cycles: 10 4 ; material: 1.4948 or 1.4919; layer thickness of folds: 0.5 mm; angular deflection (DN 800): +3 deg. C or; axial expansion absorption (DN 600): ±80 mm; calculation: ASME class. The bellows' development work is not handled within this scope. The bellows are supplied by leading manufacturers, and warrant highest quality. Multiple bellows were selected on the basis of maximum elasticity - a property

  10. LMFBR thermal-striping evaluation

    International Nuclear Information System (INIS)

    Brunings, J.E.

    1982-10-01

    Thermal striping is defined as the fluctuating temperature field that is imposed on a structure when fluid streams at different temperatures mix in the vicinity of the structure surface. Because of the uncertainty in structural damage in LMFBR structures subject to thermal striping, EPRI has funded an effort for the Rockwell International Energy Systems Group to evaluate this problem. This interim report presents the following information: (1) a Thermal Striping Program Plan which identifies areas of analytic and experimental needs and presents a program of specific tasks to define damage experienced by ordinary materials of construction and to evaluate conservatism in the existing approach; (2) a description of the Thermal Striping Test Facility and its operation; and (3) results from the preliminary phase of testing to characterize the fluid environment to be applied in subsequent thermal striping damage experiments

  11. Welding development for LMFBR applications

    International Nuclear Information System (INIS)

    Slaughter, G.M.; Edmonds, D.P.; Goodwin, G.M.; King, J.F.; Moorhead, A.J.

    1976-01-01

    High-quality welds with suitable properties for long-time elevated-temperature nuclear service are among the most critical needs in today's welding technology. Safe, reliable, and economic generation of future power depends on welded construction in systems such as Liquid Metal Fast Breeder Reactors (LMFBRs). Rapid thermal transients in LMFBR systems at coolant temperatures around 590 to 650 0 C (1000 to 1200 0 F) could cause creep and creep-fatigue damage that is not encountered in lower temperature reactor systems. The undesirable consequences of interaction between the two working fluids - sodium and steam - in the steam generators are also of major concern. Thus sound welds that have excellent reliability over a 30-year service life are essential. Several programs are actively underway at ORNL to satisfy this critical need and selected portions of three of these programs are discussed briefly

  12. Operational-safety advantages of LMFBR's: the EBR-II experience and testing program

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lindsay, R.W.; Golden, G.H.

    1982-01-01

    LMFBR's contain many inherent characteristics that simplify control and improve operating safety and reliability. The EBR-II design is such that good advantage was taken of these characteristics, resulting in a vary favorable operating history and allowing for a program of off-normal testing to further demonstrate the safe response of LMFBR's to upsets. The experience already gained, and that expected from the future testing program, will contribute to further development of design and safety criteria for LMFBR's. Inherently safe characteristics are emphasized and include natural convective flow for decay heat removal, minimal need for emergency power and a large negative reactivity feedback coefficient. These characteristics at EBR-II allow for ready application of computer diagnosis and control to demonstrate their effectiveness in response to simulated plant accidents. This latter testing objective is an important part in improvements in the man-machine interface

  13. Cesium vapor cycle for an advanced LMFBR

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    A review indicates that a cesium vapor topping cycle appears attractive for use in the intermediate fluid circuit of an advanced LMFBR designed for a reactor outlet temperature of 1250 0 F or more and would have the following advantages: (1) it would increase the thermal efficiency by about 5 to 10 points (from approximately 40 percent to approximately 45 to 50 percent) thus reducing the amount of waste heat rejected to the environment by 15 to 30 percent. (2) the higher thermal efficiency should reduce the overall capital cost of the reactor plant in dollars per kilowatt. (3) the cesium can be distilled out of the intermediate fluid circuit to leave it bone-dry, thus greatly reducing the time and cost of maintenance work (particularly for the steam generator). (4) the large volume and low pressure of the cesium vapor region in the cesium condenser-steam generator greatly reduces the magnitude of pressure fluctuations that might occur in the event of a leak in a steam generator tube, and the characteristics inherent in a condenser make it easy to design for rapid concentration of any noncondensibles that may form as a consequence of a steam leak into the cesium region so that a steam leak can be detected easily in the very early stages of its development

  14. An appreciation of the events, models and data used for LMFBR radiological source term estimations

    International Nuclear Information System (INIS)

    Keir, D.; Clough, P.N.

    1989-01-01

    In this report, the events, models and data currently available for analysis of accident source terms in liquid metal cooled fast neutron reactors are reviewed. The types of hypothetical accidents considered are the low probability, more extreme types of severe accident, involving significant degradation of the core and which may lead to the release of radionuclides. The base case reactor design considered is a commercial scale sodium pool reactor of the CDFR type. The feasibility of an integrated calculational approach to radionuclide transport and speciation (such as is used for LWR accident analysis) is explored. It is concluded that there is no fundamental obstacle, in terms of scientific data or understanding of the phenomena involved, to such an approach. However this must be regarded as a long-term goal because of the large amount of effort still required to advance development to a stage comparable with LWR studies. Particular aspects of LMFBR severe accident phenomenology which require attention are the behaviour of radionuclides during core disruptive accident bubble formation and evolution, and during the less rapid sequences of core melt under sodium. The basic requirement for improved thermal hydraulic modelling of core, coolant and structural materials, in these and other scenarios, is highlighted as fundamental to the accuracy and realism of source term estimations. The coupling of such modelling to that of radionuclide behaviour is seen as the key to future development in this area

  15. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  16. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  17. Structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    1983-01-01

    Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start

  18. Structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Vaze, M.K.K.

    1983-01-01

    The use of elastic analysis for structural design of LMFBR components is discussed. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed Prototype Fast Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is same as that of Rapsodie. Nevertheless, the design had to be checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, ASME Code Section III and the Code Case N-47 are used for high temperature design. The problems faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's shakedown and plastic cycling criteria for ratchet free operation to biaxial stress fields

  19. Status of LMFBR development project in Japan

    International Nuclear Information System (INIS)

    Nagane, G.; Akebi, M.; Matsuno, Y.

    1987-01-01

    Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

  20. Attenuation of airborne debris from LMFBR accidents

    International Nuclear Information System (INIS)

    Morewitz, H.A.; Johnson, R.P.; Nelson, C.T.; Vaughan, E.U.; Guderjahn, C.A.; Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

    1978-01-01

    Experimental and theoretical studies have been performed to characterize the behavior of airborne particulates (aerosols) expected to be produced by hypothetical core disassembly accidents (HCDA's) in liquid metal fast breeder reactors (LMFBR's). These aerosol studies include work on aerosol transport in a 20-m high, 850-m 3 closed vessel at moderate concentrations; aerosol transport in a small vessel under conditions of high concentration (approximately 1,000 g/m 3 ), high turbulence, and high temperature (approximately 2000 0 C); and aerosol transport through various leak paths. These studies have shown that tittle, if any, airborne debris from LMFBR HCDA's would reach the atmosphere exterior to an intact reactor containment building. (author)

  1. Scale modelling in LMFBR safety

    International Nuclear Information System (INIS)

    Cagliostro, D.J.; Florence, A.L.; Abrahamson, G.R.

    1979-01-01

    This paper reviews scale modelling techniques used in studying the structural response of LMFBR vessels to HCDA loads. The geometric, material, and dynamic similarity parameters are presented and identified using the methods of dimensional analysis. Complete similarity of the structural response requires that each similarity parameter be the same in the model as in the prototype. The paper then focuses on the methods, limitations, and problems of duplicating these parameters in scale models and mentions an experimental technique for verifying the scaling. Geometric similarity requires that all linear dimensions of the prototype be reduced in proportion to the ratio of a characteristic dimension of the model to that of the prototype. The overall size of the model depends on the structural detail required, the size of instrumentation, and the costs of machining and assemblying the model. Material similarity requires that the ratio of the density, bulk modulus, and constitutive relations for the structure and fluid be the same in the model as in the prototype. A practical choice of a material for the model is one with the same density and stress-strain relationship as the operating temperature. Ni-200 and water are good simulant materials for the 304 SS vessel and the liquid sodium coolant, respectively. Scaling of the strain rate sensitivity and fracture toughness of materials is very difficult, but may not be required if these effects do not influence the structural response of the reactor components. Dynamic similarity requires that the characteristic pressure of a simulant source equal that of the prototype HCDA for geometrically similar volume changes. The energy source is calibrated in the geometry and environment in which it will be used to assure that heat transfer between high temperature loading sources and the coolant simulant and that non-equilibrium effects in two-phase sources are accounted for. For the geometry and flow conitions of interest, the

  2. Active accumulation of internal DIC pools reduces transport limitation in large colonies of Nostoc pruniforme

    DEFF Research Database (Denmark)

    Raun, Ane-Marie Løvendahl; Borum, Jens; Jensen, Kaj Sand

    2009-01-01

    Nostoc pruniforme is a freshwater cyanobacterium forming large spherical colonies of up to several centimeters in diameter. The size and shape result in low surface area to volume (SA/V) ratios that potentially put severe constraints on resource acquisition. In the present study we have specifica......Nostoc pruniforme is a freshwater cyanobacterium forming large spherical colonies of up to several centimeters in diameter. The size and shape result in low surface area to volume (SA/V) ratios that potentially put severe constraints on resource acquisition. In the present study we have...

  3. Operating conditions of steam generators for LMFBR's

    International Nuclear Information System (INIS)

    Ratzel, W.

    1975-01-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  4. Operating conditions of steam generators for LMFBR's

    Energy Technology Data Exchange (ETDEWEB)

    Ratzel, W

    1975-07-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  5. THE LMFBR, key to the future

    International Nuclear Information System (INIS)

    Chipman, G.L. Jr.

    1982-01-01

    This survey explains the United States prospects for utilizing the LMFBR as a mean of meeting future energy demands. Nuclear option will represent a good financial investment only when breeder will be proved as a cost-effective option. International cooperation and combined programs are very helpful to develop breeder reactor power resource

  6. LMFBR safety experiment facility planning and analysis

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Scott, J.H.

    1976-01-01

    In the past two years considerable effort has been placed on the planning and design of new facilities for the resolution of LMFBR safety issues. The paper reviews the key issues, the experiments needed to resolve them, and the design aspects of proposed new facilities. In addition, it presents a decision theory approach to selecting an optimal combination of modified and new facilities

  7. Proposal for computer investigation of LMFBR core meltdown accidents

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

    1974-01-01

    The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

  8. Analysis of large two phase uranium dioxide bubble behavior in water and sodium pools

    International Nuclear Information System (INIS)

    Webb, R.L.

    1984-05-01

    An understanding of the behavior of large, two-phase UO 2 bubbles is important in assessing the consequences of a hypothetical core disruptive accident in a fast reactor. The UVABUBL II computer program was written to study the dynamics and heat and mass transfer in large UO 2 bubbles, and the code was used to analyze data from the underwater and undersodium FAST experiments conducted at Oak Ridge National Laboratory in which the behavior of UO 2 bubbles under a wide variety of conditions was examined. Significant understanding of the phenomena that govern UO 2 bubble behavior in both water and sodium was obtained by matching calculations of pressure, bubble size, and bubble growth and collapse rate to the experimental data. Heat and mass transfer included radiative heat losses and coolant entrainment. Larger heat transfer rates were calculated for the water tests with significant surface vaporization occurring. Because of the high thermal conductivity of sodium, no surface vaporization was calculated for the sodium tests. Entrainment was not found to be necessary for either the water or sodium tests, but calculations that included entrainment implied that it may be occurring. 38 references

  9. Bubble behavior in LMFBR core disruptive accidents. Annual report, June 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Kennedy, M.F.; Rao, S.P.; Refling, J.G.

    1976-08-01

    The work reported here is part of the Aerosol Release and Transport program for LMFBR safety assessment for the Reactor Safety Research Division of the U.S. Nuclear Regulatory Commission. Six areas were at various stages of investigation during this reporting period. A study of nonequilibrium mass transfer during fuel expansion and a study of the dynamics of fuel expansion into the sodium pool were completed. Studies are underway on condensation on above-core structures and on generation of aerosols from condensation. Studies were initiated on small-particle generation from hydrodynamic fragmentation, on particle kinematics and on particle-surface interaction

  10. Pipe supports and anchors - LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1983-06-01

    Pipe design and support design can not be treated as separate disciplines. A coordinated design approach is required if LMFBR pipe system adequacy is to be achieved at a reasonable cost. It is particularly important that system designers understand and consider those factors which influence support train flexibility and thus the pipe system dynamic stress levels. The system approach must not stop with the design phase but should continue thru the erection and acceptance test procedures. The factors that should be considered in the design of LMFBR pipe supports and anchors are described. The various pipe support train elements are described together with guidance on analysis, design and application aspects. Post erection acceptance and verification test procedures are then discussed

  11. LMFBR with booster pump in pumping loop

    International Nuclear Information System (INIS)

    Rubinstein, H.J.

    1975-01-01

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

  12. Materials engineering issues, LMFBR steam generators

    International Nuclear Information System (INIS)

    Spalaris, C.N.; Challenger, K.D.; Day, R.A.; Dutina, D.; Ring, P.J.

    1976-01-01

    Selection of 2-1/4 Cr-1 Mo as the reference construction material for LMFBR steam generators assumed a balance between its known intrinsic properties and our ability to accommodate certain of its deficiencies through design allowance. A comprehensive development program was undertaken to define base data needed, confirm assumptions made relative to desired performance, minimize defects by optimization of melting, fabrication and heat treatment processes, and prepare specifications for purchasing reactor components

  13. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    International Nuclear Information System (INIS)

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO 2 ) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO 2 pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs

  14. A critical experimental study of integral physics parameters in simulated LMFBR meltdown cores

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Wade, D.C.; Bucher, R.G.; Smith, D.M.; McKnight, R.D.; Lesage, L.G.

    1978-01-01

    Integral physics parameters of several representative, idealized meltdown LMFBR configurations were measured in mockup critical assemblies on the ZPR-9 reactor at Argonne National Laboratory. The experiments were designed to provide data for the validation of analytical methods used in the neutronics part of LMFBR accident analysis. Large core distortions were introduced in these experiments (involving 18.5% core volume) and the reactivity worths of configuration changes were determined. The neutronics parameters measured in the various configurations showed large changes upon core distortion. Both diffusion theory and transport theory methods were shown to mispredict the experimental configuration eigenvalues. In addition, diffusion theory methods were shown to result in a non-conservative misprediction of the experimental configuration change worths. (author)

  15. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  16. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo

    1976-12-01

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  17. Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

    International Nuclear Information System (INIS)

    Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

    1982-01-01

    One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

  18. Emergency air cleaning system development for LMFBR containments

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

    1975-01-01

    Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

  19. Issues in the selection of the LMFBR steam cycle

    International Nuclear Information System (INIS)

    Buschman, H.W.; McConnell, R.J.

    1983-01-01

    Unlike the light-water reactor, the liquid-metal fast breeder reactor (LMFBR) allows the designer considerable latitude in the selection of the steam cycle. This latitude in selection has been exercised by both foreign and domestic designers, and thus, despite the fact that over 25 LMFBR's have been built or are under construction, a consensus steam cycle has not yet evolved. This paper discusses the LMFBR steam cycles of interest to the LMFBR designer, reviews which of these cycles have been employed to date, discusses steam-cycle selection factors, discusses why a consensus has not evolved, and finally, concludes that the LMFBR steam-cycle selection is primarily one of technical philosophy with several options available

  20. Validation of Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water

    Directory of Open Access Journals (Sweden)

    Hua Li

    2014-01-01

    Full Text Available The Effective Heat Source (EHS and Effective Momentum Source (EMS models have been proposed to predict the development of thermal stratification and mixing during a steam injection into a large pool of water. These effective models are implemented in GOTHIC software and validated against the POOLEX STB-20 and STB-21 tests and the PPOOLEX MIX-01 test. First, the EHS model is validated against STB-20 test which shows the development of thermal stratification. Different numerical schemes and grid resolutions have been tested. A 48×114 grid with second order scheme is sufficient to capture the vertical temperature distribution in the pool. Next, the EHS and EMS models are validated against STB-21 test. Effective momentum is estimated based on the water level oscillations in the blowdown pipe. An effective momentum selected within the experimental measurement uncertainty can reproduce the mixing details. Finally, the EHS-EMS models are validated against MIX-01 test which has improved space and time resolution of temperature measurements inside the blowdown pipe. Excellent agreement in averaged pool temperature and water level in the pool between the experiment and simulation has been achieved. The development of thermal stratification in the pool is also well captured in the simulation as well as the thermal behavior of the pool during the mixing phase.

  1. Assessment of inspectability of LMFBR designs. Final report

    International Nuclear Information System (INIS)

    1981-09-01

    This two-volume report provides a comprehensive review of the inspectability of specific portions of loop- and pool-type LMFBR (1000-MWe) designs selected by EPRI. The designs were developed during the mid to late 1970s by three independent design teams (General Electric Co., Rockwell International, and Westinghouse) under the sponsorship of DOE (formerly ERDA) and EPRI. The requirements for normal, contingency, and post-repair inspections, addressed in this report, were established from Draft 12 of the ASME Boiler and Pressure Vessel Code, Section XI Division 3, issued in September 1979. These requirements, the intrinsic characteristics of the designs, the environmental (radiation, thermal, and atmospheric) aspects, and the available (present and near-term) inspection techniques, formed the basis for assessing the selected portions of the design or (1) accessibility, (2) feasibility, (3) practicality, and (4) costs to perform the above-specified inspections. Changes and additions fly ash has been as a concrete additive; however, extensive pilot scale development is underway to advance ash use in the TVA region in such areas as mineral and magnetite recovery, and mineral wool insulation. Recommended studies include: (1) the feasibility of converting existing wet fly d by the fuels include: residential (which includes residential and commercial), elthodology will be developed and verified in Phase II

  2. Effective Momentum and heat flux models for simulation of stratification and mixing in a large pool of water

    International Nuclear Information System (INIS)

    Hua Li; Villanueva, W.; Kudinov, P.

    2012-06-01

    Performance of a boiling water reactor (BWR) containment is mostly determined by reliable operation of pressure suppression pool which serves as a heat sink to cool and condense steam released from the core vessel. Thermal stratification in the pool can significantly impede the pool's pressure suppression capacity. A source of momentum is required in order to break stratification and mix the pool. It is important to have reliable prediction of transient development of stratification and mixing in the pool in different regimes of steam injection. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (1) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (2) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The POOLEX/PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized, to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale modeling, and (c) validate our proposed models. Specifically the data from POOLEX STB-21 and PPOOLEX STR-03 and STR-04 tests are used for validation of the EHS and EMS models in this work. We

  3. Effective Momentum and heat flux models for simulation of stratification and mixing in a large pool of water

    Energy Technology Data Exchange (ETDEWEB)

    Hua Li; Villanueva, W.; Kudinov, P. [Royal Institute of Technology (KTH). Div. of Nuclear Power Safety, Stockholm (Sweden)

    2012-06-15

    Performance of a boiling water reactor (BWR) containment is mostly determined by reliable operation of pressure suppression pool which serves as a heat sink to cool and condense steam released from the core vessel. Thermal stratification in the pool can significantly impede the pool's pressure suppression capacity. A source of momentum is required in order to break stratification and mix the pool. It is important to have reliable prediction of transient development of stratification and mixing in the pool in different regimes of steam injection. Previously, we have proposed to model the effect of steam injection on the mixing and stratification with the Effective Heat Source (EHS) and the Effective Momentum Source (EMS) models. The EHS model is used to provide thermal effect of steam injection on the pool, preserving heat and mass balance. The EMS model is used to simulate momentum induced by steam injection in different flow regimes. The EMS model is based on the combination of (1) synthetic jet theory, which predicts effective momentum if amplitude and frequency of flow oscillations in the pipe are given, and (2) model proposed by Aya and Nariai for prediction of the amplitude and frequency of oscillations at a given pool temperature and steam mass flux. The complete EHS/EMS models only require the steam mass flux, initial pool bulk temperature, and design-specific parameters, to predict thermal stratification and mixing in a pressure suppression pool. In this work we use EHS/EMS models implemented in containment thermal hydraulic code GOTHIC. The POOLEX/PPOOLEX experiments (Lappeenranta University of Technology, Finland) are utilized, to (a) quantify errors due to GOTHIC's physical models and numerical schemes, (b) propose necessary improvements in GOTHIC sub-grid scale modeling, and (c) validate our proposed models. Specifically the data from POOLEX STB-21 and PPOOLEX STR-03 and STR-04 tests are used for validation of the EHS and EMS models in this

  4. Psychometric characteristics of the MATRICS Consensus Cognitive Battery in a large pooled cohort of stable schizophrenia patients.

    Science.gov (United States)

    Georgiades, Anastasia; Davis, Vicki G; Atkins, Alexandra S; Khan, Anzalee; Walker, Trina W; Loebel, Antony; Haig, George; Hilt, Dana C; Dunayevich, Eduardo; Umbricht, Daniel; Sand, Michael; Keefe, Richard S E

    2017-12-01

    The MATRICS Consensus Cognitive Battery (MCCB) was developed to assess cognitive treatment effects in schizophrenia clinical trials, and is considered the FDA gold standard outcome measure for that purpose. The aim of the present study was to establish pre-treatment psychometric characteristics of the MCCB in a large pooled sample. The dataset included 2616 stable schizophrenia patients enrolled in 15 different clinical trials between 2007 and 2016 within the United States (94%) and Canada (6%). The MCCB was administered twice prior to the initiation of treatment in 1908 patients. Test-retest reliability and practice effects of the cognitive composite score, the neurocognitive composite score, which excludes the domain Social Cognition, and the subtests/domains were examined using Intra-Class Correlations (ICC) and Cohen's d. Simulated regression models explored which domains explained the greatest portion of variance in composite scores. Test-retest reliability was high (ICC=0.88) for both composite scores. Practice effects were small for the cognitive (d=0.15) and neurocognitive (d=0.17) composites. Simulated bootstrap regression analyses revealed that 3 of the 7 domains explained 86% of the variance for both composite scores. The domains that entered most frequently in the top 3 positions of the regression models were Speed of Processing, Working Memory, and Visual Learning. Findings provide definitive psychometric characteristics and a benchmark comparison for clinical trials using the MCCB. The test-retest reliability of the MCCB composite scores is considered excellent and the learning effects are small, fulfilling two of the key criteria for outcome measures in cognition clinical trials. Copyright © 2017 Elsevier B.V. All rights reserved.

  5. Operating temperatures for an LMFBR

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1993-01-01

    The scope of the present paper is limited to structural mechanics aspects that are associated with this technology. However, for the purpose of comprehensive presentation, all the other related issues are also highlighted. For this study, a Prototype Fast Breeder Reactor (PFBR) with 500 MWe capacity is taken as the reference design. Accordingly, some critical high temperature components of PFBR are analysed in- detail for elastic, inelastic and viscoplastic behaviour towards life prediction as per the requirement of design codes (RCC-MR 87) which form basis for justifying the possibility of higher operating temperatures for LMFBRs. Since operation with higher primary sodium outlet temperature in association with higher ΔT across the core is one of the efficient techniques towards making LMFBRs cost effective, operating Temperature limits are determined for a typical pool type FBR of 500 MWe capacity. Analysis indicates that control plug in the hot pool is the most critical component which limits the operating temperature to 820 K with a ΔT across the core of 160 K. By improving the thermal hydraulic design in conjunction with the structural design optimisation at the plate-shell junctions of control plug, possibility exists to go up to 840-850 K for primary outlet sodium with a T of 160 K across the core. This will result in producing steam of about 790-800 K (520 deg. C). Apart from improving the thermal hydraulic design to mitigate the transient thermal stresses, following are also needed to demonstrate higher safety margins in the design. Reduction of thermal transients, for an example, the temperature drop in the primary sodium outlet can be reduced by decreasing the sodium flow rate to the core, during a reactor scram. Welds should be avoided at the plate-shell junctions of control plug. A complete ring with necessary fillet radius may be forged as a single piece. In case of reactor vessel, a pullout option is better for redan-stand pipe junction

  6. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  7. Benchmark calculation programme concerning typical LMFBR structures

    International Nuclear Information System (INIS)

    Donea, J.; Ferrari, G.; Grossetie, J.C.; Terzaghi, A.

    1982-01-01

    This programme, which is part of a comprehensive activity aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, should allow to get confidence in computer codes which are supposed to provide a realistic prediction of the LMFBR component behaviour. The calculations started on static analysis of typical structures made of non linear materials stressed by cyclic loads. The fluid structure interaction analysis is also being considered. Reasons and details of the different benchmark calculations are described, results obtained are commented and future computational exercise indicated

  8. LMFBR accident delineation study: approach and preliminary results

    International Nuclear Information System (INIS)

    Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

    1979-01-01

    Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

  9. CEC activities in the field of LMFBR safety

    International Nuclear Information System (INIS)

    Balz, W.; Finzi, S.; Klersy, R.

    1976-01-01

    The aim of the ECC is to reach a common LMFBR Safety strategy in Europe. To this end the Commission promotes collaboration between the different fast reactor projects in the Community through working groups and collaborative arrangements and contributes with a research activity executed in its Joint Research Centre Ispra. A short description is given of the activity in the working groups and of the Ispra programme on LMFBR Safety. This programme covers: LMFBR thermohydraulics, fuel coolant interactions, dynamic structure loading and response, safety related material properties and whole core accident code development

  10. Development of a code for description of sodium spray and pool fires. Pt. 1

    International Nuclear Information System (INIS)

    Alexas, A.

    1979-09-01

    In the scope of the development of a code to describe both, sodium pool- and spray fires, the well-known codes SOFIRE II and NABRAND have been compared. Regarding the program technique of both codes, the NABRAND-code seems to be the better one, though it includes some conservatisms in the modelling and in the transport coefficients used. For a realistic estimation of the consequences of large sodium fires in an LMFBR, an elimination of these conservatisms is necessary. After that it must be investigated if a combination of the modificated version of the NABRAND-code and of a spray fire-code (for example the code SPRAY) is efficient. (orig.) [de

  11. Cash pooling

    OpenAIRE

    Lozovaya, Karina

    2009-01-01

    This work makes a mention of cash management. At next chapter describes two most known theoretical models of cash management -- Baumol Model and Miller-Orr Model. Principal part of work is about cash pooling, types of cash pooling, cash pooling at Czech Republic and influence of cash pooling over accounting and taxes.

  12. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  13. Estimation of the algal-available phosphorus pool in sediments of a large, shallow eutrophic lake (Taihu, China) using profiled SMT fractional analysis

    International Nuclear Information System (INIS)

    Zhu Mengyuan; Zhu Guangwei; Li Wei; Zhang Yunlin; Zhao Linlin; Gu Zhao

    2013-01-01

    Because large, shallow lakes are heavily influenced by wind–wave disturbance, it is difficult to estimate internal phosphorus load using traditional methods. To estimate the potential contribution of phosphorus from sediment to overlying water in eutrophic Lake Taihu, phosphorus fractions of surface and deep layer sediments were quantified and analyzed for algal bloom potential using a Standard Measurements and Testing (SMT) sequential extraction method and incubation experiments. Phosphorus bound to Fe, Al and Mn oxides and hydroxides (Fe–P) and organic phosphorus (OP) were to be found bioactive. The difference in Fe–P and OP contents between surface and deep layers equates to the sediment pool of potentially algal-available phosphorus. This pool was estimated at 5168 tons for the entire lake and was closely related to pollution input and algal blooms. Profiled SMT fractionation analysis is thus a potentially useful tool for estimating internal phosphorus loading in large, shallow lakes. - Highlights: ► We used profiled sediment P activity by SMT fractionation to evaluate the P release potential in large and shallow lakes. ► We built the relationship between sediment SMT fractionations of P and the P release by algal bloom degradation process. ► We discussed the supporting mechanism of sediment P release to Microcystis algal bloom in a large and shallow lake. ► We discussed the nutrient control strategy of algal bloom in shallow lakes in highly human activities disturbance catchment. - Profiled SMT fractional analysis of internal phosphorus pool in large, shallow lake.

  14. Simulation of droplet impact onto a deep pool for large Froude numbers in different open-source codes

    Science.gov (United States)

    Korchagova, V. N.; Kraposhin, M. V.; Marchevsky, I. K.; Smirnova, E. V.

    2017-11-01

    A droplet impact on a deep pool can induce macro-scale or micro-scale effects like a crown splash, a high-speed jet, formation of secondary droplets or thin liquid films, etc. It depends on the diameter and velocity of the droplet, liquid properties, effects of external forces and other factors that a ratio of dimensionless criteria can account for. In the present research, we considered the droplet and the pool consist of the same viscous incompressible liquid. We took surface tension into account but neglected gravity forces. We used two open-source codes (OpenFOAM and Gerris) for our computations. We review the possibility of using these codes for simulation of processes in free-surface flows that may take place after a droplet impact on the pool. Both codes simulated several modes of droplet impact. We estimated the effect of liquid properties with respect to the Reynolds number and Weber number. Numerical simulation enabled us to find boundaries between different modes of droplet impact on a deep pool and to plot corresponding mode maps. The ratio of liquid density to that of the surrounding gas induces several changes in mode maps. Increasing this density ratio suppresses the crown splash.

  15. Applicability of the Reactor Safety Study (WASH-1400) to LMFBR risk assessments

    International Nuclear Information System (INIS)

    El-Sheikh, K.A.; Feller, K.G.; Fleischer, L.; Greebler, P.; McDonald, A.; Sultan, P.; Temme, M.I.; Fullwood, R.R.

    1976-01-01

    The feasibility of applying the WASH-1400 methods and data to LMFBR risk assessment is evaluated using the following approach for a selected LMFBR: (1) Structuring the LMFBR risk assessment problem in a modular form similar to WASH-1400; (2) Comparing the predictive tools applicable to each module; (3) Comparing the dependencies among the various modules. It is concluded that the WASH-1400 applicability is limited due to LWR-LMFBR differences in operating environments and accident phenomena. WASH-1400 and LMFBR specific methods applicable to LMFBR risk assessments are indicated

  16. LMFBR subassembly response to local pressure loadings: an experimental approach

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1975-01-01

    An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

  17. Assessing the general safety and tolerability of vildagliptin: value of pooled analyses from a large safety database versus evaluation of individual studies

    Science.gov (United States)

    Schweizer, Anja; Dejager, Sylvie; Foley, James E; Kothny, Wolfgang

    2011-01-01

    Aim: Analyzing safety aspects of a drug from individual studies can lead to difficult-to-interpret results. The aim of this paper is therefore to assess the general safety and tolerability, including incidences of the most common adverse events (AEs), of vildagliptin based on a large pooled database of Phase II and III clinical trials. Methods: Safety data were pooled from 38 studies of ≥12 to ≥104 weeks’ duration. AE profiles of vildagliptin (50 mg bid; N = 6116) were evaluated relative to a pool of comparators (placebo and active comparators; N = 6210). Absolute incidence rates were calculated for all AEs, serious AEs (SAEs), discontinuations due to AEs, and deaths. Results: Overall AEs, SAEs, discontinuations due to AEs, and deaths were all reported with a similar frequency in patients receiving vildagliptin (69.1%, 8.9%, 5.7%, and 0.4%, respectively) and patients receiving comparators (69.0%, 9.0%, 6.4%, and 0.4%, respectively), whereas drug-related AEs were seen with a lower frequency in vildagliptin-treated patients (15.7% vs 21.7% with comparators). The incidences of the most commonly reported specific AEs were also similar between vildagliptin and comparators, except for increased incidences of hypoglycemia, tremor, and hyperhidrosis in the comparator group related to the use of sulfonylureas. Conclusions: The present pooled analysis shows that vildagliptin was overall well tolerated in clinical trials of up to >2 years in duration. The data further emphasize the value of a pooled analysis from a large safety database versus assessing safety and tolerability from individual studies. PMID:21415917

  18. Assessing the general safety and tolerability of vildagliptin: value of pooled analyses from a large safety database versus evaluation of individual studies

    Directory of Open Access Journals (Sweden)

    Schweizer A

    2011-02-01

    Full Text Available Anja Schweizer1, Sylvie Dejager2, James E Foley3, Wolfgang Kothny31Novartis Pharma AG, Basel, Switzerland; 2Novartis Pharma SAS, Rueil-Malmaison, France; 3Novartis Pharmaceuticals Corporation, East Hanover, NJ, USAAim: Analyzing safety aspects of a drug from individual studies can lead to difficult-to-interpret results. The aim of this paper is therefore to assess the general safety and tolerability, including incidences of the most common adverse events (AEs, of vildagliptin based on a large pooled database of Phase II and III clinical trials.Methods: Safety data were pooled from 38 studies of ≥12 to ≥104 weeks' duration. AE profiles of vildagliptin (50 mg bid; N = 6116 were evaluated relative to a pool of comparators (placebo and active comparators; N = 6210. Absolute incidence rates were calculated for all AEs, serious AEs (SAEs, discontinuations due to AEs, and deaths.Results: Overall AEs, SAEs, discontinuations due to AEs, and deaths were all reported with a similar frequency in patients receiving vildagliptin (69.1%, 8.9%, 5.7%, and 0.4%, respectively and patients receiving comparators (69.0%, 9.0%, 6.4%, and 0.4%, respectively, whereas drug-related AEs were seen with a lower frequency in vildagliptin-treated patients (15.7% vs 21.7% with comparators. The incidences of the most commonly reported specific AEs were also similar between vildagliptin and comparators, except for increased incidences of hypoglycemia, tremor, and hyperhidrosis in the comparator group related to the use of sulfonylureas.Conclusions: The present pooled analysis shows that vildagliptin was overall well tolerated in clinical trials of up to >2 years in duration. The data further emphasize the value of a pooled analysis from a large safety database versus assessing safety and tolerability from individual studies.Keywords: type 2 diabetes, dipeptidyl peptidase-4, edema, safety, vildagliptin

  19. Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

    2016-08-15

    Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident management and on eventual retrofitting measures to be implemented in the SFP building.

  20. Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Allard, L.T.; Elerath, J.G.

    1976-02-01

    This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

  1. Exciting Pools

    Science.gov (United States)

    Wright, Bradford L.

    1975-01-01

    Advocates the creation of swimming pool oscillations as part of a general investigation of mechanical oscillations. Presents the equations, procedure for deriving the slosh modes, and methods of period estimation for exciting swimming pool oscillations. (GS)

  2. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  3. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  4. Status of U.S. LMFBR programme

    International Nuclear Information System (INIS)

    Yevich, J.

    1978-01-01

    The determents of the decision for deterrence of commercial reprocessing and further demonstration of the plutonium breeder were based on two premises: time is needed to establish the programme for non-proliferating fuel cycle and there is a lessened sense of urgency for the USA to establish a commercial breeder in the near future. A strong, well funded base technology effort remains and will continue until institutional and technical solutions can be found to minimize or eliminate the proliferation risk. An LMFBR option will be maintained. The FFTF will be coming on line providing a powerful tool in breeder fuel and materials development and a baseline from which to scale up heat transfer systems and components. Sodium system hardware development and testing will continue to have high priority

  5. Work plan: transient release from LMFBR fuel

    International Nuclear Information System (INIS)

    Kress, T.S.; Parker, G.W.; Fontana, M.H.

    1975-09-01

    The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

  6. Acoustic leak detection of LMFBR steam generator

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo

    1993-01-01

    The development of a water leak detector with short response time for LMFBR steam generators is required to prevent the failure propagation caused by the sodium-water reaction and to maintain structural safety in steam generators. The development of an acoustic leak detector assuring short response time has attracted. The purpose of this paper is to confirm the basic detection feasibility of the active acoustic leak detector, and to investigate the leak detection method by erasing the background noise by spectrum analysis of the passive acoustic leak detector. From a comparison of the leak detection sensitivity of the active and the passive method, the active method is not influenced remarkably by the background noise, and it has possibility to detect microleakage with short response time. We anticipate a practical application of the active method in the future. (author)

  7. Nuclear welding, application for an LMFBR

    International Nuclear Information System (INIS)

    Patriarca, P.; Goodwin, G.M.

    1975-01-01

    Fabrication of an LMFBR system is discussed, with emphasis on areas where joint welding innovations have been introduced. Each major component of the system, including reactor vessel, intermediate heat exchanger, steam generator, and sodium-containment piping, is treated separately. Developmet of special filler metals to avoid the low elevated-temperature creep ductility obtained with conventional austenitic stainless steel weldments is reported. Bore-side welding of steam generator tube-to-tubesheet joints with and without filler metal is desirable to improve inspectability and eliminate the crevice inherent with face-side weld design, thus minimizing corrosion problems. Automated welding methods for sodium-containment piping are summarized which iminimize and control distortion and ensure welds of high integrity. Selection of materials for the various components is discussed for plants presently under construction, and materials predictions are made for future concepts. (U.S.)

  8. Microprocessor-based integrated LMFBR core surveillance

    International Nuclear Information System (INIS)

    Gmeiner, L.

    1984-06-01

    This report results from a joint study of KfK and INTERATOM. The aim of this study is to explore the advantages of microprocessors and microelectronics for a more sophisticated core surveillance, which is based on the integration of separate surveillance techniques. Due to new developments in microelectronics and related software an approach to LMFBR core surveillance can be conceived that combines a number of measurements into a more intelligent decision-making data processing system. The following techniques are considered to contribute essentially to an integrated core surveillance system: - subassembly state and thermal hydraulics performance monitoring, - temperature noise analysis, - acoustic core surveillance, - failure characterization and failure prediction based on DND- and cover gas signals, and - flux tilting techniques. Starting from a description of these techniques it is shown that by combination and correlation of these individual techniques a higher degree of cost-effectiveness, reliability and accuracy can be achieved. (orig./GL) [de

  9. Low cycle fatigue of irradiated LMFBR materials

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1976-01-01

    A review of low cycle fatigue data on irradiated LMFBR materials was conducted and extensive graphical representations of available data are presented. Representative postirradiation tensile properties of annealed 304 and 316 SS are selected and employed in several predictive methods to estimate irradiated material fatigue curves. Experimental fatigue data confirm the use of predictive methods for establishing conservative design curves over the range of service conditions relevant to such CRBRP components as core former, fixed radial shielding, core barrel, lower inlet module and upper internals structures. New experimental data on fatigue curves and creep-fatigue interaction in irradiated 20 percent cold worked (CW) 316 SS and Alloy 718 would support the design of removable radial shielding and upper internals in CRBRP. New experimental information on notched fatigue behavior and cyclic stress-strain curves of all these materials in the irradiated condition could provide significant design data

  10. Licensing decisions and safety research related to LMFBR accidents

    International Nuclear Information System (INIS)

    Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

    1977-01-01

    The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

  11. 85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

    International Nuclear Information System (INIS)

    Fair, C.E.; Cook, M.E.; Huber, K.A.; Rohde, R.

    1983-01-01

    The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system

  12. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

    International Nuclear Information System (INIS)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

  13. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  14. LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

    This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

  15. Evaluation of the LMFBR cover gas source term and synthesis of the associated R and D

    International Nuclear Information System (INIS)

    Balard, F.; Carluec, B.

    1996-01-01

    At the end of the seventies and the beginning of the eighties, there appeared a pressing need of experimental results to assess the LMFBR's safety level. Because of the urgency, analytical studies were not systematically undertaken and maximum credible cover gas instantaneous source terms (radionuclides core release fraction) were got directly from crude out-of-pile experiment interpretations. Two types of studies and mock-ups were undertaken depending on the timescale of the phenomena: instantaneous source terms (corresponding to an unlikely energetic core disruptive accident CDA), and delayed ones (tens of minutes to some hours). The experiments performed in this frame are reviewed in this presentation: 1) instantaneous source term: - FAUST experiments: I, Cs, UO2 source terms (FzK, Germany), - FAST experiments : pool depth influence on non volatile source term (USA), - CARAVELLE experiments: nonvolatile source term in SPX1 geometry (CEA, France); 2) delayed source term: - NALA experiments: I, Cs, Sr, UO2 source term (FzK, Germany), - PAVE experiments: I source term (CEA, France), - NACOWA experiments: cover gas aerosols enrichment in I and Cs (FzK, Germany) - other French experiments in COPACABANA and GULLIVER facilities. The volatile fission products release is tightly bound to sodium evaporation and a large part of the fission products is dissolved in the liquid sodium aerosols present in the cover gas. Thus the knowledge of the amount of aerosol release to the cover gas is important for the evaluation of the source term. The maximum credible cover gas instantaneous source terms deduced from the experiments have led to conservative source terms to be taken into account in safety analysis. Nevertheless modelling attempts of the observed (in-pile or out-of-pile) physico-chemical phenomena have been undertaken for extrapolation to the reactor case. The main topics of this theoretical research are as follows: fission products evaporation in the cover gas (Fz

  16. Vitamin D Pooling Project

    Science.gov (United States)

    The Vitamin D Pooling Project of Rarer Cancers brought together investigators from 10 cohorts to conduct a large prospective epidemiologic study of the association between vitamin D status and seven rarer cancers.

  17. Sodium pool fire model for CONACS code

    International Nuclear Information System (INIS)

    Yung, S.C.

    1982-01-01

    The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA

  18. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

    Simmons, W.R.

    1984-01-01

    The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

  19. Hydrogen formation and control under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Armstrong, G.R.; Wierman, R.W.

    1976-09-01

    The objective of this study is to experimentally investigate the potential for autoignition and combustion of hydrogen-sodium mixtures which may be produced in LMFBR accidents. The purpose and ultimate usefulness of this work is to provide data that will establish the validity and acceptability of mechanisms inherent to the LMFBR that could either prevent or delay the accumulation of hydrogen gas to less than 4 percent (V) in the Reactor Containment Building (RCB) under accident conditions. The results to date indicate that sodium and sodium-hydrogen mixtures such as may be expected during LMFBR postulated accidents will ignite upon entering an air atmosphere and that the hydrogen present will be essentially all consumed until such time that the oxygen concentration is depleted

  20. Cost-competitive, inherently safe LFMBR pool plant

    International Nuclear Information System (INIS)

    McDonald, J.S.; Brunings, J.E.; Chang, Y.I.; Hren, R.R.; Seidensticker, R.W.

    1984-01-01

    The Cost-Competitive, Inherently Safe LMFBR Pool Plant design was prepared in GFY 1983 under a DOE-sponsored program. This plant design was developed as a joint effort by Rockwell International and the Argonne National Laboratory with major contributions from the Bechtel Group, Inc.; Combustion engineering, Inc.; the Chicago Bridge and Iron Company; and the General Electric Company. Using current LMFBR technology, many innovative features were developed and incorporated into the design to meet the ultimate objectives of the Breeder Program, i.e., energy costs competitive with LWRs and inherent safety features to maintain the plant in a safe condition following assumed accidents without requiring operator action. This paper provides a description of the principal features that were incorporated into the design to achieve low cost and inherent safety

  1. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  2. Computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

    1977-01-01

    Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

  3. Coolant mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-06-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

  4. Review of PRA methodology for LMFBR

    International Nuclear Information System (INIS)

    Yang, J. E.

    1999-02-01

    Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs

  5. Intelligent type sodium instrumentations for LMFBR

    International Nuclear Information System (INIS)

    Chen Daolong

    1996-07-01

    The constructions and performances of lots of newly developed intelligent type sodium instrumentations are described. The graduation characteristic equations for corresponding transducer using the medium temperature as the parameter are given. These intelligent type sodium instrumentations are possessed of good linearity. The accurate measurement data of sodium process parameters (flowrate, pressure and level) can be obtained by means of their on-line compensation function of the temperature effect. Moreover, these intelligent type sodium instrumentations are possessed of the self-inspection, the electric shutoff protection, the setting of full-scale, the setting of alarm limits (two upper limits and two lower limits alarms), the thermocouple breaking alarm, mutual isolative the 0∼10 V direct-current analogue output and the CENTRONICS standard digital output, and the alarm relay contact output. Theses intelligent type sodium instrumentations are suitable particularly for the instrument, control and protective systems of LMFBR by means of these excellent functions based on microprocessor. The basic errors of the intelligent type sodium flowmeter, immersed sodium flowmeter, sodium manometer and sodium level gauge are +-2%, +-2.3%, +-0.3% and +-1.9% of measuring ranges respectively. (9 figs.)

  6. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  7. 54Mn release from LMFBR cores

    International Nuclear Information System (INIS)

    Polley, M.V.

    1976-10-01

    The inventory of 54 Mn per unit exposed area of stainless steel in LMFBR cores may be calculated using a formula originally derived at HEDL. This treats the simultaneous production by activation and release by corrosion and diffusion of 54 Mn and assumes that the concentration at the steel surface is zero. The inventory per unit exposed area is calculated as a function of temperature and is compared with that calculated simply by assuming stoichiometric corrosion. An effective diffusion coefficient is used in the calculations which include contributions from both lattice and grain boundary diffusion. A general relationship is derived for the effective diffusion coefficient and it is shown how values may be obtained using the Levine-MacCallum and the Fisher theories of grain boundary diffusion. Values of the lattice diffusion coefficient were obtained by analysing data obtained from sodium loop experiments. The effect on the inventory due to the possible formation of a ferrite layers on the exposed surface is discussed and it is also shown how the inventory over several fuel cycles may be calculated. (U.K.)

  8. Diets high in resistant starch and arabinoxylan modulate digestion processes and SCFA pool size in the large intestine and faecal microbial composition in pigs.

    Science.gov (United States)

    Nielsen, Tina S; Lærke, Helle N; Theil, Peter K; Sørensen, Jens F; Saarinen, Markku; Forssten, Sofia; Knudsen, Knud E Bach

    2014-12-14

    The effects of a high level of dietary fibre (DF) either as arabinoxylan (AX) or resistant starch (RS) on digestion processes, SCFA concentration and pool size in various intestinal segments and on the microbial composition in the faeces were studied in a model experiment with pigs. A total of thirty female pigs (body weight 63.1 (sem 4.4) kg) were fed a low-DF, high-fat Western-style control diet (WSD), an AX-rich diet (AXD) or a RS-rich diet (RSD) for 3 weeks. Diet significantly affected the digestibility of DM, protein, fat, NSP and NSP components, and the arabinose:xylose ratio, as well as the disappearance of NSP and AX in the large intestine. RS was mainly digested in the caecum. AX was digested at a slower rate than RS. The digesta from AXD-fed pigs passed from the ileum to the distal colon more than twice as fast as those from WSD-fed pigs, with those from RSD-fed pigs being intermediate (PEubacterium rectale, Bifidobacterium spp. and Lactobacillus spp. in the faeces sampled at week 3 of the experimental period (P< 0.05). In the caecum, proximal and mid colon, AXD feeding resulted in a 3- to 5-fold higher pool size of butyrate compared with WSD feeding, with the RSD being intermediate (P <0.001). In conclusion, the RSD and AXD differently affected digestion processes compared with the WSD, and the AXD most efficiently shifted the microbial composition towards butyrogenic species in the faeces and increased the large-intestinal butyrate pool size.

  9. Measurements of dynamic shape factors of LMFBR aggregate aerosols

    International Nuclear Information System (INIS)

    Allen, M.D.; Moss, O.R.; Briant, J.K.

    1980-01-01

    Dynamic shape factors for branched, chain-like aggregates of LMFBR mixed-oxide fuels have been measured with a LAPS spiral-duct centrifuge. The aerosol was generated by repeatedly pulsing a focused laser beam onto the surface of a typical LMFBR fuel pellet. The measured values of the dynamic shape factor, corrected for slip, vary between kappa = 3.60 at D/sub ae/ = 0.5 μm, and kappa = 2.23 at D/sub ae/ = 1.5 μm

  10. Airborne effluent control for LMFBR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-01-01

    A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including 131 I, krypton, tritium, 129 I, and most recently 14 C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems

  11. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  12. Status of gamma-ray heating characterization in LMFBR

    International Nuclear Information System (INIS)

    Gold, R.

    1975-11-01

    Efforts to define gamma-ray heating in Liquid Metal Fast Breeder Reactor (LMFBR) environments have been surveyed. Emphasis is placed on both current practice for the Experimental Breeder Reactor-II (EBR-II) and future needs of the Fast Flux Test Facility (FFTF). Experimental and theoretical work are included in this preliminary survey for both high and low power environments. Current ''state-of-the-art'' accuracies and limitations are assessed. On this basis, it is concluded that a broad and sustained effort be initiated to meet requested FFTF goal accuracies. To this end, recommendations are advanced for improving the current status of gamma heating characterization and temperature measurements in LMFBR

  13. Neutronic characteristics simulation of LMFBR of great size

    International Nuclear Information System (INIS)

    Kim, Y.C.

    1987-09-01

    The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr

  14. PDA: Pooled DNA analyzer

    Directory of Open Access Journals (Sweden)

    Lin Chin-Yu

    2006-04-01

    Full Text Available Abstract Background Association mapping using abundant single nucleotide polymorphisms is a powerful tool for identifying disease susceptibility genes for complex traits and exploring possible genetic diversity. Genotyping large numbers of SNPs individually is performed routinely but is cost prohibitive for large-scale genetic studies. DNA pooling is a reliable and cost-saving alternative genotyping method. However, no software has been developed for complete pooled-DNA analyses, including data standardization, allele frequency estimation, and single/multipoint DNA pooling association tests. This motivated the development of the software, 'PDA' (Pooled DNA Analyzer, to analyze pooled DNA data. Results We develop the software, PDA, for the analysis of pooled-DNA data. PDA is originally implemented with the MATLAB® language, but it can also be executed on a Windows system without installing the MATLAB®. PDA provides estimates of the coefficient of preferential amplification and allele frequency. PDA considers an extended single-point association test, which can compare allele frequencies between two DNA pools constructed under different experimental conditions. Moreover, PDA also provides novel chromosome-wide multipoint association tests based on p-value combinations and a sliding-window concept. This new multipoint testing procedure overcomes a computational bottleneck of conventional haplotype-oriented multipoint methods in DNA pooling analyses and can handle data sets having a large pool size and/or large numbers of polymorphic markers. All of the PDA functions are illustrated in the four bona fide examples. Conclusion PDA is simple to operate and does not require that users have a strong statistical background. The software is available at http://www.ibms.sinica.edu.tw/%7Ecsjfann/first%20flow/pda.htm.

  15. From Swimming Pool to Collaborative Learning Studio: Pedagogy, Space, and Technology in a Large Active Learning Classroom

    Science.gov (United States)

    Lee, Dabae; Morrone, Anastasia S.; Siering, Greg

    2018-01-01

    To promote student learning and bolster student success, higher education institutions are increasingly creating large active learning classrooms to replace traditional lecture halls. Although there have been many efforts to examine the effects of those classrooms on learning outcomes, there is paucity of research that can inform the design and…

  16. Feasibility study for adapting ITREC plant to reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Moccia, A.; Rolandi, G.

    1976-05-01

    The report evaluates the feasibility of adapting ITREC plant to the reprocessing LMFBR fuels, with the double purpose of: 1) recovering valuable Pu contained in these fuels and recycling it to the fabrication plant; 2) trying, on a pilot scale, the chemical process technology to be applied in a future industrial plant for reprocessing the fuel elements discharged from fast breeder power reactors

  17. Refractory metal carbide coatings for LMFBR applications: a systems approach

    International Nuclear Information System (INIS)

    Gotschall, H.L.; Ople, F.S.; Riccardella, P.C.

    1975-01-01

    The selection, testing and improvement of high density, tightly bonded plasma and detonation gun coatings designed to meet LMFBR core component criteria are described. The process descriptions include a review of the important developments in substrate surface preparation which were required to ensure strong bonding and to minimize interface contamination. Coating finishing techniques which were developed to optimize friction behavior are also described

  18. German position paper on structural analysis for LMFBR applications

    International Nuclear Information System (INIS)

    Angerbauer, A.; Link, F.

    1983-01-01

    During the design period of the German LMFBR, the SNR-300, extensive work had been done in the field of elastic and inelastic analysis. Furthermore, special design rules have been developed. A review of these activities and their state-of-the art is outlined in this paper

  19. Fission-gas bubble modeling for LMFBR accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1977-01-01

    The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

  20. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  1. Small leak shutdown, location, and behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Sandusky, D.W.

    1976-01-01

    The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

  2. Studies of LMFBR: method of analysis and some results

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.; Nascimento, J.A. do.

    1983-01-01

    Some results of recent studies of LMFBR characteristics are summarized. A two-dimensional model of the LMFBR is taken from a publication and used as the base model for the analysis. Axial structures are added to the base model and a three-dimensional (Δ - Z) calculation has been done. Two dimensional (Δ and RZ) calculations are compared with the three-dimensional and published results. The eigenvalue, flux and power distributions, breeding characteristics, control rod worth, sodium-void and Doppler reactivities are analysed. Calculations are done by CITATION using six-group cross sections collapsed regionwise by EXPANDA in one-dimensional geometries from the 70-group JFS library. Burnup calculations of a simplified thorium-cycle LMFBR have also been done in the RZ geometry. Principal results of the studies are: (1) the JFS library appears adequate for predicting overall characteristics of an LMFBR, (2) the sodium void reactivity is negative within - 25 cm from the outer boundary of the core, (3) the halflife of Pa-233 must be considered explicitly in burnup analyses, and (4) two-dimensional (RZ and Δ) calculations can be used iteratively to analyze three-dimensional reactor systems. (Author) [pt

  3. Corrosion critique of the 2 1/4 Cr--1 Mo steel for LMFBR steam generation system applications

    International Nuclear Information System (INIS)

    Zima, G.E.

    1977-07-01

    The unstabilized ferritic steel of nominal composition, 2 1 / 4 Cr-1Mo, has been proposed for critical structural assignments in LMFBR powerplants, specifically: the tubing, tubesheet and shell of the evaporator and superheater components. The interest in this steel has been based on a presumably favorable general corrosion property spectrum, acceptable mechanical properties and fabricability, and certain economies associated with the low alloy content. This report is an attempt at a general corrosion assessment for the 2 1 / 4 Cr-1Mo steel and an identification of corrosion problem areas potential to this steel from the sodium and water/steam systems of the proposed working environment. There is a considerable area of uncertainty in the sodium-side response of 2 1 / 4 Cr-1Mo steel, centered in the loss and redisposition of carbon during long-term exposure to sodium of various impurity backgrounds. It is submitted that present evidence relating to the water/steam-side corrosion behavior of the 2 1 / 4 Cr-1Mo steel, under nominal and conceivable perturbed environmental conditions, constitutes the principal concern for the proposed LMFBR powerplant applications of this steel. It is suggested that this unfavorable corrosion aspect represents an inherent limitation of the low alloy content of this steel, probably largely independent of melting and processing recourses, and it is a sufficient basis to question the incentive for a continuation of the collateral studies of this steel for the proposed LMFBR steam generation system assignments

  4. Environmental genomics of "Haloquadratum walsbyi" in a saltern crystallizer indicates a large pool of accessory genes in an otherwise coherent species

    Directory of Open Access Journals (Sweden)

    Bolhuis Henk

    2006-07-01

    Full Text Available Abstract Background Mature saturated brine (crystallizers communities are largely dominated (>80% of cells by the square halophilic archaeon "Haloquadratum walsbyi". The recent cultivation of the strain HBSQ001 and thesequencing of its genome allows comparison with the metagenome of this taxonomically simplified environment. Similar studies carried out in other extreme environments have revealed very little diversity in gene content among the cell lineages present. Results The metagenome of the microbial community of a crystallizer pond has been analyzed by end sequencing a 2000 clone fosmid library and comparing the sequences obtained with the genome sequence of "Haloquadratum walsbyi". The genome of the sequenced strain was retrieved nearly complete within this environmental DNA library. However, many ORF's that could be ascribed to the "Haloquadratum" metapopulation by common genome characteristics or scaffolding to the strain genome were not present in the specific sequenced isolate. Particularly, three regions of the sequenced genome were associated with multiple rearrangements and the presence of different genes from the metapopulation. Many transposition and phage related genes were found within this pool which, together with the associated atypical GC content in these areas, supports lateral gene transfer mediated by these elements as the most probable genetic cause of this variability. Additionally, these sequences were highly enriched in putative regulatory and signal transduction functions. Conclusion These results point to a large pan-genome (total gene repertoire of the genus/species even in this highly specialized extremophile and at a single geographic location. The extensive gene repertoire is what might be expected of a population that exploits a diverse nutrient pool, resulting from the degradation of biomass produced at lower salinities.

  5. Overview of current activities relevant to structural analysis on LMFBR in Japan

    International Nuclear Information System (INIS)

    Ichimiya, Masakazu

    1983-01-01

    This paper presents the structural analysis activities on LMFBR in Japan. The structural analysis activities on LMFBR in Japan have been made mainly toward the validation of the rules of high temperature structural design guide which is to be used for the design of Class 1 components for elevated temperature service of the prototype fast breeder reactor, Monju. Main features of these analyses are as follows. (1) Since the design by elastic analysis is intended in the high temperature structural design guide of Monju, a large progress has been made in the bounding technique for high temperature inelastic behaviors, particularly the elastic follow-up. (2) There has been a progress in the clarification of the creep behavior in order to evaluate creep damage adequately. (3) Analysis techniques and design rules for piping have been developed with considerable emphasis. In addition, buckling analyses were performed considering the thin structures with low internal pressure in Monju components. Further test and analysis were made on ratcheting. (author)

  6. Large-scale precipitation tracking and the MJO over the Maritime Continent and Indo-Pacific warm pool

    Science.gov (United States)

    Kerns, Brandon W.; Chen, Shuyi S.

    2016-08-01

    A large-scale precipitation tracking (LPT) method is developed to track convection and precipitation associated with the Madden-Julian oscillation (MJO) using the Tropical Rainfall Measurement Mission 3B42 rainfall data from October to March 1998-2015. LPT uses spatially smoothed 3 day rainfall accumulation to identify and track precipitation features in time with a minimum size of 300,000 km2 and time continuity at least 10 days. While not all LPT systems (LPTs) are attributable to the MJO, among the 199 LPTs, there were 42 with a mean eastward propagation of at least 2 m s-1, which are considered to be MJO convective initiation events. These LPTs capture the diversity of the MJO convection, which is not well depicted by the Real-time Multivariate MJO (RMM) index or the outgoing longwave radiation MJO index. During the 17 years, there were 17 instances out of 45 with a MJO signature in the RMM without eastward propagating LPTs. Among the 42 eastward propagating LPTs, 24 propagated across the Maritime Continent (MC), which confirms the MC barrier effect. Among the cases that crossed the MC from the Indian Ocean to the western Pacific (MC crossing), 18 (75%) had a significant MJO signature in the RMM index. In contrast, only six (33%) of the non-MC-crossing cases occurred with a RMM MJO signal. There is a significant seasonal and interannual variability with MC-crossing LPTs occurring in December more commonly than other months. More MC-crossing events were observed during La Niña than El Niño, which is consistent with the observations of stronger and more frequent MJO events identified by RMM during La Niña years.

  7. Experimental plans for LMFBR cavity liner sodium spill test LT-1

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Newell, G.A.

    1976-01-01

    Reinforced concrete is an important material of construction in LMFBR cavities and cells. Steel liners are often installed on the concrete surfaces to provide a gastight seal for minimizing air inleakage to inerted cell atmospheres and to protect the concrete from direct contact with sodium in the event of a sodium spill. In making safety assessment analyses, it is of interest to determine the adequacy of the liners to maintain their leaktightness during postulated accidents involving large sodium spills. However, data for basing analytical assessments of cell liners are very meager and an experimental program is underway at HEDL to provide some of the needed information. The HEDL cell liner evaluation program consists of both bench-scale feature tests and large-scale sodium spill demonstration tests. The plans for the first large-scale sodium spill test (LT-1) are the subject of this paper

  8. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

    Anderson, R.; Hayden, O.

    2002-01-01

    Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

  9. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  10. Application of 2-1/4 Cr-1 Mo as a structural material in saturated steam cycle LMFBR systems. Final report

    International Nuclear Information System (INIS)

    Licina, G.J.; Busboom, H.J.; Ring, P.J.; Roy, P.; Schmidt, C.G.; Spalaris, C.N.

    1982-02-01

    The suitability and incentives were examined for using 2-1/4Cr-1Mo steel as a structural material for the entire primary and secondary sodium systems in a 1000 MWe pool-type Liquid Metal Fast Breeder Reactor. The critical properties, advantages and disadvantages of 2-1/4Cr-1Mo, and data needed for design were described for each major component in the reactor. The relative importance of alloy properties to the successful use of ferritics in LMFBR was identified. Licensing issues, likely to surface if ferritic alloys were to be used for critical reactor components, were discussed

  11. Water tests for determining post voiding behavior in the LMFBR

    International Nuclear Information System (INIS)

    Hinkle, W.D.

    1976-06-01

    The most serious of the postulated accidents considered in the design of the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) is the Loss of Pipe Integrity (LOPI) accident. Analysis models used to calculate the consequences of this accident assume that once boiling is initiated film dryout occurs in the hot assembly as a result of rapid vapor bubble growth and consequent flow stoppage or reversal. However, this assumption has not been put to any real test. Once boiling is initiated in the hot assembly during an LMFBR LOPI accident, a substantial gravity pressure difference would exist between this assembly and other colder assemblies in the core. This condition would give rise to natural circulation flow boiling accompanied by pressure and flow oscillations. It is possible that such oscillations could prevent or delay dryout and provide substantial post-voiding heat removal. The tests described were conceived with the objective of obtaining basic information and data relating to this possibility

  12. Hydrogen jet recombination under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Wierman, R.W.

    1977-01-01

    Certain conditions may be postulated in LMFBR risk assessments for which the potential of hydrogen release to the reactor containment building needs to be evaluated. The inherent self-ignition characteristics of hydrogen jets entering the air atmosphere of the reactor containment building should be understood for such analyses. If hydrogen jets were to self-ignite (recombine) at the source where they enter the reactor containment building, then undesirable hydrogen accumulation would not occur. Therefore, experiments have been conducted investigating the phenomena associated with the recombination of hydrogen jets under conditions similar to those postulated for LMFBR studies. The data presented define the conditions required for self-ignition of the hydrogen jets

  13. Design and economic implications of heterogeneity in an LMFBR core

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1983-01-01

    Much emphasis is currently being placed in LMFBR design on reducing both the capital cost and the fuel cycle cost of an LMFBR to insure its economic competativeness without a rapid increase in the uranium prices. In this study the relationship between two core design options, their neutronic consequences, and their effect on fuel cycle cost are analyzed. The two design options are the selection of pin diameter and the degree of heterogeneity. In the case of a heterogeneous core, with a low sodium void reactivity worth this ratio of fertile internal blanket to driver assemblies is generally about 0.40. However, some advantages of cores with heterogeneity of 0.08 to 0.2 for a fixed pin diameter have been reported

  14. Pool scrubbing

    International Nuclear Information System (INIS)

    Lopez-Jimenez, J.; Herranz, J.; Escudero, M.J.; Espigares, M.M.; Peyres, V.; Polo, J.; Kortz, Ch.; Koch, M.K.; Brockmeier, U.; Unger, H.; Dutton, L.M.C.; Smedley, Ch.; Trow, W.; Jones, A.V.; Bonanni, E.; Calvo, M.; Alonso, A.

    1996-12-01

    The Source Term Project in the Third Frame Work Programme of the European Union Was conducted under and important joined effort on pool scrubbing research. CIEMAT was the Task Manager of the project and several other organizations participated in it: JRC-Ispra, NNC Limited, RUB-NES and UPM. The project was divided into several tasks. A peer review of the models in the pool scrubbing codes SPARC90 and BUSCA-AUG92 was made, considering the different aspects in the hydrodynamic phenomenology, particle retention and fission product vapor abortions. Several dominant risk accident sequences were analyzed with MAAP, SPARC90 and BUSCA-AUG92 codes, and the predictions were compared. A churn-turbulent model was developed for the hydrodynamic behaviour of the pool. Finally, an experimental programme in the PECA facility of CIEMAT was conducted in order to study the decontamination factor under jet injection regime, and the experimental observations were compared with the SPARC and BUSCA codes. (Author)

  15. Cover-gas seals: 11-LMFBR seal-test program

    International Nuclear Information System (INIS)

    Steele, O.P. III; Horton, P.H.

    1977-01-01

    The objective of the Cover Gas Seal Material Development Program is to perform the engineering development required to provide reliable seals for LMFBR application. Specific objectives are to verify the performance of commercial solid cross-section and inflatable seals under reactor environments including radiation, to develop advanced materials and configurations capable of achieving significant improvement in radioactive gas containment and seal temperature capabilities, and to optimize seal geometry for maximum reliability and minimal gas permeation

  16. Axial migratin of cesium in LMFBR fuel pins

    International Nuclear Information System (INIS)

    Karnesky, R.A.; Bridges, A.E.; Jost, J.W.

    1981-11-01

    A correlated model for quantitatively predicting the behavior of cesium in LMFBR fuel pins has been developed. This correlation was shown to be in good agreement with experimental data. It has been used to predict the behavior of cesium in the FFTF driver fuel and as the result of this analysis it has been shown that the accumulation of cesium in the insulator pellets at the ends of the fuel column will not be life limiting

  17. Retention of gaseous fission products in reprocessing LMFBR fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-05-01

    The report is devoted to status of the development programme at the Oak Ridge National Laboratory on methods for retaining iodine-131 and 129, Krypton-85, Tritium and Carbon-14 in reprocessing LMFBR fuels. The Iodox process, Fluorocarbon absorption process and Voloxidation process are described for retention of iodine, Krypton-85 and Tritium, respectively. Flowsheets for the different processes are given and results of experimental runs in small engineering-scale equipment are reported

  18. Development of acidic processes for decontaminating LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    Hill, E F [Rockwell International, Atomics International Division, Canoga Park (United States); Colburn, R P; Lutton, J M; Maffei, H P [Hanford Engineering Development Laboratory, Richland (United States)

    1978-08-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components.

  19. LMFBR operational safety: the EBR-II experience

    International Nuclear Information System (INIS)

    Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

    1978-01-01

    The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

  20. Design approaches to achieve competitive LMFBR capital costs

    International Nuclear Information System (INIS)

    Arnold, W.H.; Ehrman, C.S.; Sharbaugh, J.E.; Young, W.H.

    1982-01-01

    Through analysis of the essential functional elements of an LMFBR, numerous ways were found to simplify system design, reduce the size of components and equipment, and eliminate some components and systems. The projected capital cost per net kW of this design is competitive with that of current PWRs. RandD programs and the construction and operation of CRBRP now are needed to prove out the features of this new design

  1. Thermal analysis methods for LMFBR wire wrapped bundles

    International Nuclear Information System (INIS)

    Todreas, N.E.

    1976-11-01

    A note is presented which was written to stimulate an awareness and discussion of the fundamental differences in the formulation of certain existing analysis codes for LMFBR wire wrap bundles. The contention of the note is that for those array types where data exists (one wire per pin, equal start angles), the ENERGY method results for coolant temperature under forced convection conditions provide benchmarks of reliability equal to the results of codes COBRA and TH1-3D

  2. Structural analysis for elevated temperature design of the LMFBR

    International Nuclear Information System (INIS)

    Griffin, D.S.

    1976-02-01

    In the structural design of LMFBR components for elevated temperature service it is necessary to take account of the time-dependent, creep behavior of materials. The accommodation of creep to assure design reliability has required (1) development of new design limits and criteria, (2) development of more detailed representations of material behavior, and (3) application of the most advanced analysis techniques. These developments are summarized and examples are given to illustrate the current state of technology in elevated temperature design

  3. Analytical work on local faults in LMFBR subassembly

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Miyaguchi, K.; Hirata, N.; Kasahara, F.

    1979-01-01

    Analytical codes have been developed for evaluating various severe but highly unlikely events of local faults in the LMFBR subassembly (S/A). These include: (1) local flow blockage, (2) two-phase thermohydraulics under fission gas release, and (3) inter-S/A failure propagation. A simple inter-S/A thermal failure propagation analysis code, FUMES, is described that allows an easy parametric study of propagation potential of fuel fog in a S/A. 7 refs

  4. A new approach to the design of LMFBR liners

    International Nuclear Information System (INIS)

    Polentz, L.M.

    1980-01-01

    An advance in the state-of-the-art of LMFBR liners which permits notable savings in construction costs without any sacrifice of safety is described. The application of the new design concept to the rework of the upper reactor vault liner of the FFTF is discussed. Factors which affect the application of the new design approach to other LMFBRs are delineated and discussed. (author)

  5. Effect of operating temperature on LMFBR core performance

    International Nuclear Information System (INIS)

    Noyes, R.C.; Bergeron, R.J.; di Lauro, G.F.; Kulwich, M.R.; Stuteville, D.W.

    1977-01-01

    The purpose of the study is to provide an engineering evaluation of high and low temperature LMFBR core designs. The study was conducted by C-E supported by HEDL expertise in the areas of materials behavior, fuel performance and fabrication/fuel cycle cost. The evaluation is based primarily on designs and analyses prepared by AI, GE and WARD during Phase I of the PLBR studies

  6. Route survey for LMFBR spent fuel transportation analysis

    International Nuclear Information System (INIS)

    Foley, J.T.

    1977-05-01

    Descriptions are given of surveys that were made along segments of interstate highways to obtain information on objects near the right-of-ways and on highway features that constitute hazards in the event of transportation accidents. Data collected during the surveys are summarized. The work was done in support of the LMFBR Hazards Analysis which was being performed for the Division of Reactor Development and Demonstration of the U.S. Energy Research and Development Administration

  7. LMFBR technology. FFTF cover-gas leakage calculation

    International Nuclear Information System (INIS)

    Deboi, H.

    1974-01-01

    The FFTF LMFBR is intended to have a near zero release of radioactive gases during normal reactor operation with 1% failed fuel. This report presents calculations which provide an approximation of these cover gas leakages. Data from ongoing static and dynamic seal leak tests at AI are utilized. Leakage through both elastomeric and metallic seals in all sub-assemblies and penetrations comprising the reactor cover gas containment during reactor operation system are included

  8. Development of acidic processes for decontaminating LMFBR components

    International Nuclear Information System (INIS)

    Hill, E.F.; Colburn, R.P.; Lutton, J.M.; Maffei, H.P.

    1978-01-01

    The objective of the DOE decontamination program is to develop a well characterized chemical decontamination process for application to LMFBR primary system components that subsequently permits contact maintenance and allows requalification of the components for reuse in reactors. The paper describes the subtasks of deposit characterization, development of requalification and process acceptance criteria, development of process evaluation techniques and studies which led to a new acidic process for decontaminating 304 stainless steel hot leg components

  9. Biological behavior of mixed LMFBR-fuel-sodium aerosols

    International Nuclear Information System (INIS)

    Mahlum, D.D.; Hackett, P.L.; Hess, J.O.; Allen, M.D.

    1979-01-01

    Immediately after exposure of rats to mixed aerosols of sodium-LMFBR fuel, about 80 to 90% of the body burden of 239 Pu is in the gastrointestinal tract; 1.5 to 4% is in the lungs. With fuel-only aerosols, less of the body burden was in the GI tract and more in the lung and the head. Blood and urine values suggest an increased absorption of 239 Pu from sodium-fuel than from fuel-only aerosols

  10. Transport-diffusion comparisons for small core LMFBR disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1977-11-01

    A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

  11. Pooled Resequencing of 122 Ulcerative Colitis Genes in a Large Dutch Cohort Suggests Population-Specific Associations of Rare Variants in MUC2.

    Science.gov (United States)

    Visschedijk, Marijn C; Alberts, Rudi; Mucha, Soren; Deelen, Patrick; de Jong, Dirk J; Pierik, Marieke; Spekhorst, Lieke M; Imhann, Floris; van der Meulen-de Jong, Andrea E; van der Woude, C Janneke; van Bodegraven, Adriaan A; Oldenburg, Bas; Löwenberg, Mark; Dijkstra, Gerard; Ellinghaus, David; Schreiber, Stefan; Wijmenga, Cisca; Rivas, Manuel A; Franke, Andre; van Diemen, Cleo C; Weersma, Rinse K

    2016-01-01

    Genome-wide association studies have revealed several common genetic risk variants for ulcerative colitis (UC). However, little is known about the contribution of rare, large effect genetic variants to UC susceptibility. In this study, we performed a deep targeted re-sequencing of 122 genes in Dutch UC patients in order to investigate the contribution of rare variants to the genetic susceptibility to UC. The selection of genes consists of 111 established human UC susceptibility genes and 11 genes that lead to spontaneous colitis when knocked-out in mice. In addition, we sequenced the promoter regions of 45 genes where known variants exert cis-eQTL-effects. Targeted pooled re-sequencing was performed on DNA of 790 Dutch UC cases. The Genome of the Netherlands project provided sequence data of 500 healthy controls. After quality control and prioritization based on allele frequency and pathogenicity probability, follow-up genotyping of 171 rare variants was performed on 1021 Dutch UC cases and 1166 Dutch controls. Single-variant association and gene-based analyses identified an association of rare variants in the MUC2 gene with UC. The associated variants in the Dutch population could not be replicated in a German replication cohort (1026 UC cases, 3532 controls). In conclusion, this study has identified a putative role for MUC2 on UC susceptibility in the Dutch population and suggests a population-specific contribution of rare variants to UC.

  12. Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

    International Nuclear Information System (INIS)

    Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

    1985-04-01

    The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

  13. Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.; Gvildys, J.

    1977-01-01

    In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs

  14. LMFBR post accident heat removal testing needs and conceptual design of a test facility

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

    1977-03-01

    A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

  15. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  16. Diets high in resistent starch and arabinoxylan modulate digestion processes and SCFA pool size in the large intestine and faecal microbial composition in pigs

    DEFF Research Database (Denmark)

    Nielsen, Tina Skau; Lærke, Helle Nygaard; Theil, Peter Kappel

    2014-01-01

    The effects of a high level of dietary fibre (DF) either as arabinoxylan (AX) or resistant starch (RS) on digestion processes, SCFA concentration and pool size in various intestinal segments and on the microbial composition in the faeces were studied in a model experiment with pigs. A total...... resulted in a 3- to 5-fold higher pool size of butyrate compared with WSD feeding, with the RSD being intermediate (P microbial composition towards butyrogenic...

  17. Clinical impact of tumor location on the colon cancer survival and recurrence: analyses of pooled data from three large phase III randomized clinical trials.

    Science.gov (United States)

    Aoyama, Toru; Kashiwabara, Kosuke; Oba, Koji; Honda, Michitaka; Sadahiro, Sotaro; Hamada, Chikuma; Maeda, Hiromichi; Mayanagi, Shuhei; Kanda, Mitsuro; Sakamoto, Junichi; Saji, Shigetoyo; Yoshikawa, Takaki

    2017-11-01

    The aim of the present study was to determine whether or not the overall survival (OS) and disease-free survival (DFS) were affected by the tumor location in patients who underwent curative resection for colon cancer in a pooled analysis of three large phase III studies performed in Japan. In total, 4029 patients were included in the present study. Patients were classified as having right-side colon cancer (RC) if the primary tumor was located in the cecum, ascending colon, hepatic flexure or transverse colon, and left-side colon cancer (LCC) if the tumor site was within the splenic flexure, descending colon, sigmoid colon or recto sigmoid junction. The risk factors for the OS and DFS were analyzed. In the present study, 1449 patients were RC, and 2580 were LCC. The OS rates at 3 and 5 years after surgery were 87.6% and 81.6% in the RC group and 91.5% and 84.5% in the LCC group, respectively. Uni- and multivariate analyses showed that RRC increased the risk of death by 19.7% (adjusted hazard ratio = 1.197; 95% confidence interval, 1.020-1.408; P = 0.0272). In contrast, the DFS was similar between the two locations. The present study confirmed that the tumor location was a risk factor for the OS in patients who underwent curative treatment for colon cancer. Tumor location may, therefore, need to be considered a stratification factor in future phase III trials of colon cancer. © 2017 The Authors. Cancer Medicine published by John Wiley & Sons Ltd.

  18. Clinical outcome after high-precision radiotherapy for skull base meningiomas: Pooled data from three large German centers for radiation oncology.

    Science.gov (United States)

    Combs, Stephanie E; Farzin, Mostafa; Boehmer, Julia; Oehlke, Oliver; Molls, Michael; Debus, Jürgen; Grosu, Anca-Ligia

    2018-05-01

    To evaluate outcome in patients with base of skull meningiomas treated with modern high precision radiation therapy (RT) techniques. 927 patients from three centers were treated with either radiosurgery or fractionated high-precision RT for meningiomas. Treatment planning was based on CT and MRI following institutional guidelines. For radiosurgery, a median dose of 13 Gy was applied, for fractionated treatments, a median dose of 54 Gy in 1.8 Gy single fractions was prescribed. Follow-up included a clinical examination as well as contrast-enhanced imaging. All patients were followed up prospectively after radiotherapy in the three departments within a strict follow-up regimen. The median follow-up time was 81 months (range 1-348 months). Median local control was 79 months (range 1-348 months). Local control (LC) was 98% at 1 year, 94% at 3 years, 92% at 5 years and 86% at 10 years. There was no difference between radiosurgery and fractionated RT. We analyzed the influence of higher doses on LC and could show that dose did not impact LC. Moreover, there was no difference between 54 Gy and 57.6 Gy in the fractionated group. Side effects were below 5% in both groups without any severe treatment-related complications. Based on the pooled data analysis this manuscript provides a large series of meningiomas of the skull base treated with modern high precision RT demonstrating excellent local control and low rates of side effects. Such data support the recommendation of RT for skull base meningiomas in the interdisciplinary tumor board discussions. The strong role of RT must influence treatment recommendations keeping in mind the individual risk-benefit profile of treatment alternatives. Copyright © 2018 Elsevier B.V. All rights reserved.

  19. Post-accident fuel relocation and heat removal in the LMFBR

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

  20. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  1. Volume-heated boiling pool behavior and application to transition phase accident conditions

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

    1978-01-01

    Observations of two-phase flow fields in volume-heated boiling pools are reported. Photographic observations, together with pool-average void fraction measurements are presented. Flow regime transition criterial derived from the measurements are discussed. The churn-turbulent flow regime was the dominant regime for superficial vapor velocity. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. The results of the experiment and analysis are extrapolated to transition phase conditions. It is shown that intense pool boil-up could occur where the pool-average void fraction would be greater than 0.6 for steel vaporization rates equivalent to power levels greater than one percent of nominal LMFBR power density. (author)

  2. Volume-heated boiling pool flow behavior and application to transition phase accident conditions

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

    1978-01-01

    Observations of two-phase flow fields in volume-heated boiling pools are reported. Photographic observations, together with pool-average void fraction measurements are presented. Flow regime transition criteria derived from the measurements are discussed. The churn-turbulent flow regime was the dominant regime for superficial vapor velocities up to nearly five times the Kutateladze dispersal velocity. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. The results of the experiment and analyses are extrapolated to transition phase conditions. It is shown that intense pool boil-up could occur where the pool-average void fraction would be greater than 0.6 for steel vaporization rates equivalent to power levels greater than one percent of nominal LMFBR power density

  3. LMFBR models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

  4. Evaluation of integrally finned cladding for LMFBR fuel pins

    International Nuclear Information System (INIS)

    Cantley, D.A.; Sutherland, W.H.

    1975-01-01

    An integral fin design effectively reduces the coolant temperature gradients within an LMFBR subassembly by redistributing coolant flow so as to reduce the maximum cladding temperature and increase the duct wall temperature. The reduced cladding temperatures are offset by strain concentrations resulting from the fin geometry, so there is little net effect on predicted fuel pin performance. The increased duct wall temperatures, however, significantly reduce the duct design lifetime so that the final conclusion is that the integral fin design is inferior to the standard wire wrap design. This result, however, is dependent upon the material correlations used. Advanced alloys with improved irradiation properties could alter this conclusion

  5. Fission and corrosion products behavior in primary circuits of LMFBR's

    International Nuclear Information System (INIS)

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  6. Confirmatory simulation of safety and operational transients in LMFBR systems

    International Nuclear Information System (INIS)

    Guppy, J.G.; Agrawal, A.K.

    1978-01-01

    Operational and safety transients that may originate anywhere in an LMFBR system must be adequately simulated to assist in safety evaluation and plant design efforts. This paper describes an advanced thermohydraulic transient code, the Super System Code (SSC), that may be used for confirmatory safety evaluations of plant wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions, and presents results obtained with SSC illustrating the degree of modelling detail present in the code as well as the computing efficiency. (author)

  7. Sodium water reaction R and D for French LMFBR

    International Nuclear Information System (INIS)

    Cambillard, E.; Finck, P.; Lapicore, A.; Simeon, C.

    1985-01-01

    This paper presents the research and development which is underway for the French LMFBR steam generator safety study. The program comprises three major areas: (1) the analysis of realistic leaks, which includes the leak evolution and its consequences; (2) the response time of leak detection systems compared to leak propagation phenomena; and (3) the guillotine rupture (DBA) studies relative to source term evaluation by experimental/calculational approach and mechanical calculations. This program has provided information for the demonstrations of the steam generator safety in respect to a sodium-water reaction

  8. LMFBR steam generator leak detection development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Magee, P M; Gerrels, E E; Greene, D A [General Electric Company, Sunnyvale, CA (United States); McKee, J [Argonne National Laboratory, Argonne, IL (United States)

    1978-10-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H{sub 2} and O{sub 2}) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  9. Immersed acoustical transducers and their potential uses in LMFBR

    International Nuclear Information System (INIS)

    Argous, J.P.; Brunet, M.; Baron, J.; Lhuillier, C.; Segui, J.L.

    1980-04-01

    Six years satisfactory operation in PHENIX has proved the reliability and effectivness of under-sodium viewing (VISUS) and Acoustic Detection. This fact has been strong incentive to maintain, on the future LMFBR the visus as well as the Acoustic Detection functions. These two functions are performed on SUPER PHENIX, by two sets of distinct systems using the well-known solution. Taking into account of recent improvements in sodium immersible acoustic transducers technology, CEA decided to undertake the development of a multi-functions instrument. This paper gives an outline of this new concept, which should be able to reduce the cost and the complexity of core instrumentation

  10. CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

    1983-01-01

    The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

  11. LMFBR steam generator leak detection development in the United States

    International Nuclear Information System (INIS)

    Magee, P.M.; Gerrels, E.E.; Greene, D.A.; McKee, J.

    1978-01-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H 2 and O 2 ) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  12. LMFBR fuel cycle studies progress report, August 1972, No. 42

    International Nuclear Information System (INIS)

    Unger, W.E.; Blanco, R.E.; Crouse, D.J.; Irvine, A.R.; Watson, C.D.

    1972-10-01

    This report continues a series outlining progress in the development of methods for reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO 2 --UO 2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports

  13. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  14. Malaria diagnosis from pooled blood samples: comparative analysis of real-time PCR, nested PCR and immunoassay as a platform for the molecular and serological diagnosis of malaria on a large-scale

    Directory of Open Access Journals (Sweden)

    Giselle FMC Lima

    2011-09-01

    Full Text Available Malaria diagnoses has traditionally been made using thick blood smears, but more sensitive and faster techniques are required to process large numbers of samples in clinical and epidemiological studies and in blood donor screening. Here, we evaluated molecular and serological tools to build a screening platform for pooled samples aimed at reducing both the time and the cost of these diagnoses. Positive and negative samples were analysed in individual and pooled experiments using real-time polymerase chain reaction (PCR, nested PCR and an immunochromatographic test. For the individual tests, 46/49 samples were positive by real-time PCR, 46/49 were positive by nested PCR and 32/46 were positive by immunochromatographic test. For the assays performed using pooled samples, 13/15 samples were positive by real-time PCR and nested PCR and 11/15 were positive by immunochromatographic test. These molecular methods demonstrated sensitivity and specificity for both the individual and pooled samples. Due to the advantages of the real-time PCR, such as the fast processing and the closed system, this method should be indicated as the first choice for use in large-scale diagnosis and the nested PCR should be used for species differentiation. However, additional field isolates should be tested to confirm the results achieved using cultured parasites and the serological test should only be adopted as a complementary method for malaria diagnosis.

  15. Analytical approach for confirming the achievement of LMFBR reliability goals

    International Nuclear Information System (INIS)

    Ingram, G.E.; Elerath, J.G.; Wood, A.P.

    1981-01-01

    The approach, recommended by GE-ARSD, for confirming the achievement of LMFBR reliability goals relies upon a comprehensive understanding of the physical and operational characteristics of the system and the environments to which the system will be subjected during its operational life. This kind of understanding is required for an approach based on system hardware testing or analyses, as recommended in this report. However, for a system as complex and expensive as the LMFBR, an approach which relies primarily on system hardware testing would be prohibitive both in cost and time to obtain the required system reliability test information. By using an analytical approach, results of tests (reliability and functional) at a low level within the specific system of interest, as well as results from other similar systems can be used to form the data base for confirming the achievement of the system reliability goals. This data, along with information relating to the design characteristics and operating environments of the specific system, will be used in the assessment of the system's reliability

  16. Confirmatory simulation of safety and operational transients in LMFBR systems

    International Nuclear Information System (INIS)

    Guppy, J.G.; Agrawal, A.K.

    1978-01-01

    Operational and safety transients (anticipated, unlikely, or extremely unlikely) that may originate anywhere in a liquid-metal fast breeder reactor (LMFBR) system must be adequately simulated to assist in safety evaluation and plant design efforts. An advanced thermohydraulic transient code, the Super System Code (SSC), is described that may be used for confirmatory safety evaluations of plant-wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions. Results obtained with SSC illustrating the degree of modeling detail present in the code as well as the computing efficiency are presented. A version of the SSC code, SSC-L, applicable to any loop-type LMFBR design, has been developed at Brookhaven. The scope of SSC-L is to enable the simulation of all plant-wide transients covered by Plant Protection System (PPS) action, including sodium pipe rupture and coastdown to natural circulation conditions. The computations are stopped when loss of core integrity (i.e., clad melting temperature exceeded) is indicated

  17. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  18. LMFBR core flowering response to an impulse load

    International Nuclear Information System (INIS)

    Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

    1993-01-01

    Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

  19. LMFBR subassembly response to simulated local pressure loadings

    International Nuclear Information System (INIS)

    Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

    1976-01-01

    The structural response of liquid metal fast breeder reactor (LMFBR) subassemblies to local accidental events is of interest in assessing the safety of such systems. Problems to be resolved include failure propagation modes from pin to pin and from subassembly to subassembly. Factors which must be considered include: (a) the geometry of the structure, (b) uncertainty of the pressure-energy source, (c) uncertainty of materials properties under reactor operating conditions, and (d) the difficulty in performing in-pile or out-of-pile experiments which would simulate the above conditions. The main effort in evaluating the subassembly response has been centered around the development of appropriate analyses based on the finite element technique. Analysis has been extended to include not only the subassembly duct structure itself, but also the fluid environment, both within subassemblies and between them. These models and codes have been devised to cover a wide range of accident loading conditions, and can treat various materials as their properties become known. The effort described here is centered mainly around an experimental effort aimed at verfying, modifying or extending the models used in treating subassembly damage propagation. To verify the finite element codes under development, a series of out-of-pile room temperature experiments has been performed on LMFBR-type subassembly ducts under various loading conditions. (Auth.)

  20. Transient analysis of LMFBR reinforced/prestressed concrete containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Belytschko, T.B.; Bazant, Z.P.

    1979-01-01

    The use of prestressed concrete reactor vessels (PCRVs) for LMFBR containment creates a need for analytical methods for treating the transient response of such structures, for LMFBR containments must be capable of sustaining the dynamic effects which arise in a hypothetical core disruptive accident (HCDA). These analyses require several unique features: a model of concrete which includes tensile cracking, a methodology for representing the prestressing tendons and for simulating the prestressing operation, and an efficient computational tool for treating the transient response. Furthermore, for the sake of convenience, all of these features should be available in a single computer code. For the purpose of treating the transient response, a finite element program with explicit time integration was chosen. The use of explicit time integration has the advantage that it can easily treat the complicated constitutive model which arises from the considerations of concrete cracking and it can handle the slip between reinforcing tendons and the concrete through the use of the well known sliding interface options. However, explicit time integration programs are usually not well suited to the simulation of static processes such as prestressing. Nevertheless, explicit time integration programs can handle static processes through the introduction of damping by what is known as a dynamic relaxation procedure. For this reason, the dynamic relaxation procedure was refined through the introduction of lumped mass, viscous damping. This provision made the prestressing operation of the concrete structures by means of the explicit formulation rather convenient. (orig.)

  1. LMFBR intermediate-heat-exchanger experience

    International Nuclear Information System (INIS)

    Cho, S.M.; Beaver, T.R.

    1983-01-01

    This paper presents developmental and operating experience of large Intermediate Heat Exchangers (IHX's) in US from the Fast Flux Test Facility (FFTF) to the Clinch River Breeder Reactor Plant (CRBRP) to the Large Development Plant (LDP). Design commonalities and deviations among these IHX's are synopsized. Various developmental tests that were conducted in the areas of hydraulic, structural and mechanical design are also presented. The FFTF is currently operating. Performance data of the FFTF IHXs are reviewed, and comparisons between actual and predicted performances are made. The results are used to assess the adequacy of IHX designs

  2. Hawaii ESI: POOLS (Anchialine Pool Points)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for anchialine pools in Hawaii. Anchialine pools are small, relatively shallow coastal ponds that occur...

  3. State of the art review of degradation processes in LMFBR materials. Volume II. Corrosion behavior

    International Nuclear Information System (INIS)

    Dillon, R.D.

    1975-01-01

    Degradation of materials exposed to Na in LMFBR service is reviewed. The degradation processes are discussed in sections on corrosion and mass transfer, erosion, wear and self welding, sodium--water reactions, and external corrosion. (JRD)

  4. Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

    International Nuclear Information System (INIS)

    Moreira, M.L.

    1985-01-01

    A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

  5. Seismic response and damping tests of small bore LMFBR piping and supports

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.; Lindquist, M.R.

    1984-01-01

    Seismic testing and analysis of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described. Measured responses to simulated seismic excitations are compared with analytical predictions based on NRC Regulatory Guide 1.61 and measured system damping values. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps

  6. LMFBR fuel analysis. Task A: oxide fuel dynamics. Final report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Dhir, V.K.; Frank, M.; Kastenberg, W.E.; McKone, T.E.

    1979-03-01

    Three aspects of LMFBR safety are discussed. The first concerns the potential reactivity effects of whole core fuel motion prior to pin failure in low ramp rate transient overpower accidents. The second concerns the effects of flow blockages following pin failure on the coolability of a core following an unprotected overpower transient. The third aspect concerns the safety related implications of using thorium based fuels in LMFBR's

  7. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    Han, J.T.; Fontana, M.H.

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions

  8. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  9. Are LMFBR permits unconstitutional. [German Feferal Republic

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, H; Ziegler, E

    1977-12-01

    The August 18, 1977 decision by the Muenster Higher Administrative Court to have the Federal Constitutional Court investigate the constitutionality of permits granted for fast breeder power plants has aroused much attention, both in the FRG and in other countries. This is the first time that the German Atomic Energy Act is being questioned with respect to the separation of powers between legislative and executive authorities and also with respect to the principle of a constitutional state. As a result of their analysis of the first information available about the court decision the authors have some doubts as to whether the views held by the court about the consequences of the development of fast breeder reactors and of the permit granted for the SNR 300 demonstration nuclear power station are essentially correct. In view of the wording of and the official comments on the incriminated Section 7 of the Atomic Energy Act and the large number of subsequent leading decisions by the Federal Diet about fast breeder reactors also the concern about the constitutionality of that reactor line appears to be unfounded.

  10. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

    Arnaud, G.; Guigon, A.; Verset, L.

    1983-06-01

    Since the last ten years, the French program has been, first of all, directed to the setting up, and then the development of, at once, the Phenix control rods, and next, the Super-Phenix ones. The vented pin design, with porous plug and sodium bonding, which allows the choices of large diameters, has been taken, since the Rapsodie experience was decisive. The absorber material is sintered, 10 B enriched, boron carbide. The can is made of 316 type stainless steel, stabilised, or not, with titanium. The experience gained in Phenix up to now is important, and deals with about six loads of control rods. Results confirm the validity of the design of the absorber pins. Some difficulties has been encountered for the guiding devices, due to the swelling of the steel. They have required design and material improvements. Such difficulties are discarded by a new design of the bearing, for the Super-Phenix control rods. The other parts of these rods, from the Primary Shut-Down System, are strictly derived from Phenix. The design of the rods from the Secondary Shut-Down System is rather different, but it's not the case for the design of the absorber pins: in many a way, they are derived from Phenix pins and from Rapsodie control rods. Both types of rods irradiation tests are in progress in Phenix [fr

  11. Detailed inelastic analysis of an LMFBR pipeline

    International Nuclear Information System (INIS)

    Hibbitt, H.D.; Leung, E.K.; Ohalla, A.K.

    1982-01-01

    The paper describes detailed inelastic analyses of a large diameter, thin walled pipeline configuration typical of liquid metal cooled reactor primary piping, subject to thermal shock, with intermediate periods of creep hold time. Three such analyses are compared. Two of these analyses are performed with recently developed elements based on a combination of Fourier and polynomial interpolation to describe the deformation of the pipe. One of these two analyses includes continuous deformation of the pipe wall between each elbow and the adjacent straight pipe segments, while the other neglects such ''end effects'' on the elbow deformation. The third analysis is based on a modified axi-symmetric shell element for modeling the elbows (neglecting and effects). The results thus provide an assessment of the relative cost and importance of including consideration of end effects in modeling a realistic piping system, as well as providing a similar comparison between the two basic deforming section pipe models (Fourier/polynomial versus modified axi-symmetric shells)

  12. Development of a simple estimation tool for LMFBR construction cost

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Kinoshita, Izumi

    1999-01-01

    A simple tool for estimating the construction costs of liquid-metal-cooled fast breeder reactors (LMFBRs), 'Simple Cost' was developed in this study. Simple Cost is based on a new estimation formula that can reduce the amount of design data required to estimate construction costs. Consequently, Simple cost can be used to estimate the construction costs of innovative LMFBR concepts for which detailed design has not been carried out. The results of test calculation show that Simple Cost provides cost estimations equivalent to those obtained with conventional methods within the range of plant power from 325 to 1500 MWe. Sensitivity analyses for typical design parameters were conducted using Simple Cost. The effects of four major parameters - reactor vessel diameter, core outlet temperature, sodium handling area and number of secondary loops - on the construction costs of LMFBRs were evaluated quantitatively. The results show that the reduction of sodium handling area is particularly effective in reducing construction costs. (author)

  13. MIT LMFBR blanket research project. Final summary report

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1983-08-01

    This is a final summary report on an experimental and analytical program for the investigation of LMFBR blanket characteristics carried out at MIT in the period 1969 to 1983. During this span of time, work was carried out on a wide range of subtasks, ranging from neutronic and photonic measurements in mockups of blankets using the Blanket Test Facility at the MIT Research Reactor, to analytic/numerical investigations of blanket design and economics. The main function of this report is to serve as a resource document which will permit ready reference to the more detailed topical reports and theses issued over the years on the various aspects of project activities. In addition, one aspect of work completed during the final year of the project, on doubly-heterogeneous blanket configurations, is documented for the record

  14. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  15. A technique for computing bowing reactivity feedback in LMFBR's

    International Nuclear Information System (INIS)

    Finck, P.J.

    1987-01-01

    During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

  16. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Betten, P.R.; Tow, D.M.

    1984-04-01

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  17. Validation of turbulence models for LMFBR outlet plenum flows

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-01-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds number (Re) values of 33000 and 70000 in a 1/15 - scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different two-equation turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet flow field, importantly also upon the degree of inlet turbulence, and also upon the turbulent momentum exchange model used in the calculations. In the FFTF geometry, the TEACH-T predictions agree well with the experiments. 7 refs

  18. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, T.; Mukai, K.; Yamamoto, K.

    1980-01-01

    Bellows are employed as useful mechanical elements with their flexibility and imperviousness to liquid and gas in the system in which such chemically active substances as sodium are handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: control rod drive mechanism; intermediate heat exchanger; small valve; mechanical penetration assembly of the containment boundary; outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for FBR use in Japan. (author)

  19. Research and development of bellows for LMFBR in Japan

    International Nuclear Information System (INIS)

    Takahashi, Tadao; Mukai, Kazuo; Yamamoto, Ken.

    1979-11-01

    The bellows is employed as a useful mechanical element with its flexibility and imperviousness to liquid and gas in the system in which such chemically active substance as sodium is handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: - control rod drive mechanism, - intermediate heat exchanger, - small valve, - mechanical penetration assembly of the containment boundary, - outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for the FBR use in Japan. (author)

  20. Transient behaviour and inherent safety research of LMFBR power plants

    International Nuclear Information System (INIS)

    Zhu Jizhou; Wang Ping; Yu Baoan

    1995-06-01

    Fast Breeder Reactor will be the next generation reactor for nuclear electricity production, the development of FBR will give the profits of efficient utilization of nuclear resources. The fast reactor safety analysis is the foundation and key of FBR research work. Therefore, a block-oriented mathematical model for the primary system of LMFBRs was constructed, and the dynamic simulating results which have been carried out on micro-computer are presented for various transients, i.e. TOP, LOFS, LOHS. The results agree well with the corresponding results of the code NATDEMO and experiment results of EBR-II. Based on previous analysis, various methods are discussed to confirm the inherent safety of LMFBR

  1. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  2. Fatigue of LMFBR piping due to flow stratification

    Energy Technology Data Exchange (ETDEWEB)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface.

  3. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

  4. Simple LMFBR axial-flow friction-factor correlation

    International Nuclear Information System (INIS)

    Chan, Y.N.; Todreas, N.E.

    1981-09-01

    Complicated LMFBR axial lead-length averaged friction factor correlations are reduced to an easy, ready-to-use function of bundle Reyonlds number for wire-wrapped bundles. The function together with the power curves to calculate the associated constants are incorporated in a computer pre-processor, EZFRIC. The constants required for the calculation of the subchannels and bundle friction factors are derived and correlated into power curves of geometrical parameters. A computer program, FRIC, which can alternatively be used to accurately calculate these constants is also included. The accuracte values of the constants and the corresponding values predicted by the power curves and percentage error of prediction are tabulated for a wide variety of geometries of interest

  5. Fatigue of LMFBR piping due to flow stratification

    International Nuclear Information System (INIS)

    Woodward, W.S.

    1983-01-01

    Flow stratification due to reverse flow was simulated in a 1/5-scale water model of a LMFBR primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. The behavior of the interface was characterized in terms of location, local temperature fluctuation and duration for various reverse flow conditions. A structural assessment was performed to determine the effects of stratified flow on the fatigue life of the pipe. Both the static and dynamic aspects of flow stratification were examined. The dynamic interface produces thermal striping on the inside of the pipe wall which is shown to have the most deleterious effect on the pipe wall and produce significant fatigue damage relative to a static interface

  6. Gravitational agglomeration of post-HCDA LMFBR aerosols: nonspherical particles

    International Nuclear Information System (INIS)

    Tuttle, R.F.; Loyalka, S.K.

    1982-12-01

    Aerosol behavior analysis computer programs have shown that temporal aerosol size distributions in nuclear reactor containments are sensitive to shape factors. This research investigates shape factors by a detailed theoretical analysis of hydrodynamic interactions between a nonspherical particle and a spherical particle undergoing gravitational collisions in an LMFBR environment. First, basic definitions and expressions for settling speeds and collisional efficiencies of nonspherical particles are developed. These are then related to corresponding quantities for spherical particles through shape factors. Using volume equivalent diameter as the defining length in the gravitational collision kernel, the aerodynamic shape factor, the density correction factor, and the gravitational collision shape factor, are introduced to describe the collision kernel for collisions between aerosol agglomerates. The Navier-Stokes equation in oblate spheroidal coordinates is solved to model a nonspherical particle and then the dynamic equations for two particle motions are developed. A computer program (NGCEFF) is constructed, and the dynamical equations are solved by Gear's method

  7. Creep strain accumulation in a typical LMFBR piperun

    International Nuclear Information System (INIS)

    Johnstone, T.L.

    1975-01-01

    The analysis described allows the strain concentrations in typical LMFBR two anchor point uniplanar piperuns to be calculated. Account is taken of the effect of pipe elbows in attracting creep strain to themselves as well as possible movements of the thrust line due to strain redistribution. The influence of the initial load conditions is also examined. The stress relaxation analysis is facilitated by making the assumption that a cross-sectional stress distribution determined by the asymptotic fully developed state of creep exists at all times. Use is then made of Hoff(s) analogy between materials with a creep law of the Norton type and those with a corresponding non-linear elastic stress strain law, to determine complementary strain energy rates for straight pipes and bends. Ovalisation of the latter produces an increased strain energy rate which can be simply calculated by comparison with an equal length of straight pipe through employing a creep flexibility factor due to Spence. Deflection rates at any location in the pipework can then be evaluated in terms of the thermal restraint forces at that location by an application of Castigliano's principle. In particular for an anchor point the deflection rates are identically zero and this leads to the generation of 3 simultaneous differential equations determining the relaxation of the anchor reactions. Indicative results are presented for the continuous relaxation at 570 deg C of the thermally induced stress in a planar approximation to a typical LMFBR pipe run chosen to have peak elbow stresses close to the code maximum. The results indicate a ratio, after 10 5 hours, of 3 for creep strain concentration relative to initial peak strain (calculated on the assumption of fully elastic behavior) in the most severely affected elbow, when either austenitic 316 or 321 creep properties are employed

  8. Specialists meeting on LMFBR flow induced vibrations. Summary report

    International Nuclear Information System (INIS)

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs

  9. Specialists meeting on LMFBR flow induced vibrations. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-01

    A Specialists' Meeting on LMFBR Flow-Induced Vibrations was held at ANL in the United States which was sponsored by the International Atomic Energy Agency (IAEA) on the recommendations of the International Working Group on Fast Reactors (IWGFR). It was attended by participants from France, the Federal Republic of Germany, Italy, Japan, Netherlands, the United Kingdom, the Union of Soviet Socialist Republics, the United States and the IAEA. The purpose of the meeting was to provide, for the first time, a common forum for the exchange of information on flow-induced vibration programs of the member countries. As this was a first meeting, information was sought in the broad areas of: 1. Design Criteria and Problem Areas in LMFBR Design; 2. Current Design Procedures; and 3. Ongoing Research. A session was devoted to each of the above topics wherein papers were presented and discussed followed by open discussions on the session topic. The objective of the open discussions was to identify, from a review of specific reactor designs, (a) flow induced vibration problem areas (expected and observed) and their potential for occurrence; (b) failure modes and associated design criteria; (c) specific components that are susceptible to flow induced vibration; and (d) probable excitation mechanisms. It was aimed to assess the current state-of-the-art in designing to avoid flow induced vibration with consideration of licensing requirements; to evaluate existing methods of analysis, testing, and surveillance, along with their limitations and to identify areas requiring research and review ongoing research programmes relative to these research needs.

  10. Study on the phenomena of natural circulation in LMFBR

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Koga, Tomonari

    1993-01-01

    Decay heat removal with natural circulation is to be introduced to the LMFBR operation under loss of the electric power supply. The natural circulation is highly reliable, but the phenomenon is essentially unstable and subtle, which makes fine prediction difficult. The difficulties of experimental prediction are explained by facts that the phenomena are ruled by the delicate balance between the buoyancy force and the low pressure loss and are influenced by the various parameters such as local geometry, heat capacity and so on. Therefore the similarity rule for the natural circulation has not been fully understood. This study has been conducted to establish the simulation method for the natural circulation phenomena and the detailed phenomena have been reviewed. For the natural circulation in an LMFBR plant, there are no readily available reference velocity and temperature. These values are related only with the heating and cooling rate, the characteristic length and physical properties of the testing fluid. Basic equations were transformed by these values, and dimensionless equations were derived and then two dimensionless numbers, the Gr' number and the Bo' number, were identified. In order to examine the similarity rule for natural circulation we performed experiments using the different scale water models, a 1/20th and a 1/6th model. The temperatures and velocities at typical points were measured in the transient condition with various heating rate as a parameter. Measured temperatures and velocities were transformed to dimensionless forms for comparison and the effects of the Bo' number and the Gr' number were examined. As a result, it was clarified that the effect of the Gr' number is negligibly small but the effect of Bo' number still remained in our experimental range. The Bo' number of an actual plant is within the range of this experiment. Accordingly similitude of the Bo' number becomes important in an experiment to simulate an actual plant. (author)

  11. Safety issues for LMFBR: important features drawn from the assessments of Superphenix

    International Nuclear Information System (INIS)

    Natta, M.

    2002-01-01

    Superphenix, which is built on the site of Creys-Malville, is still the biggest LMFBR plant that has been in operation. It is a pool type reactor, as Phenix and the RNR 1 500 and EFR projects. After the analysis of the preliminary safety (1974-1975), the construction was authorised by decree of the Prime Minister in 1977, the authorization for fuel loading and star-up to 3% was given by the minister of industry in July 1985 and full power was achieved in December 1986. The plant was operated until the end of December 1996, producing the equivalent of 320 EFPD, corresponding to half of the maximum barn-up of the first core. The plant was definitively stopped on the 20. of April 1998 by a decision of the French government. During this period of 25 years of licensing, construction and operation of Superphenix, others discussions and preliminary licensing procedures were started for new projects, mainly the RNR 1500 French project and the EFR European project. The operation of Superphenix was also marked by several incidents, which led to additional licensing procedures and important modifications. This period was also marked by an important work of research and development in the safety field, mostly related to the issues concerning hypothetical core disruptive accidents (HCDA) and sodium fires; further, this period was marked by the Three Mile Island accident in 1979 and the Chernobyl accident in 1986. The purpose of this paper is to present some items which were discussed during this period of 25 years and which should be of interest for future LMFBRs. In this presentation, we shall discuss the key issues concerning the safety criteria and options taken with respect to severe accidents, i.e. core melt accidents, giving details on some specific which are less known since they were assessed only lately for Superphenix, sometimes in connection with the on-going safety researches. (author)

  12. Pooling and correlated neural activity

    Directory of Open Access Journals (Sweden)

    Robert Rosenbaum

    2010-04-01

    Full Text Available Correlations between spike trains can strongly modulate neuronal activity and affect the ability of neurons to encode information. Neurons integrate inputs from thousands of afferents. Similarly, a number of experimental techniques are designed to record pooled cell activity. We review and generalize a number of previous results that show how correlations between cells in a population can be amplified and distorted in signals that reflect their collective activity. The structure of the underlying neuronal response can significantly impact correlations between such pooled signals. Therefore care needs to be taken when interpreting pooled recordings, or modeling networks of cells that receive inputs from large presynaptic populations. We also show that the frequently observed runaway synchrony in feedforward chains is primarily due to the pooling of correlated inputs.

  13. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    International Nuclear Information System (INIS)

    Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Shin, Y.S.

    1977-01-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  14. A review of ANL base technology studies in support of the U.S. LMFBR vibration program

    Energy Technology Data Exchange (ETDEWEB)

    Wambsganss, M W; Chen, S S [Components Technology Division, Argonne National Laboratory, Argonne, IL (United States); Mulcahy, T M; Shin, Y S

    1977-12-01

    Argonne National Laboratory (ANL) is the center for base technology studies of flow induced vibration for the U.S. LMFBR Program. This paper reviews and summarizes published results, reports on the status of ongoing programs, and discusses future needs as outlined in the U.S. LMFBR Vibrations Program Plan. (author)

  15. Analysis of closed-pool boilup using the TRANSIT-HYDRO code

    International Nuclear Information System (INIS)

    Graff, D.L.

    1983-01-01

    The benign termination of the transition phase of a hypothetical LMFBR accident rests on the avoidance of highly energetic recriticalities prior to escape of bottled molten core materials from the active core region. In scenarios where molten fuel is trapped due to axial blockages, the maintenance of subcritical configurations until radial flow paths develop requires stable boil-up of the molten fuel/steel mixture. This paper describes the analysis of an experiment investigating the behavior of closed boiling pools using the two-fluid hydrodynamics module of TRANSIT-HYDRO, a deterministic transition-phase analysis code

  16. Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

    International Nuclear Information System (INIS)

    Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

    1988-01-01

    This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

  17. Neutronic feasibility of an LMFBR super long-life core (SLLC)

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Aoki, Katsutada; Arie, Kazuo; Tsuboi, Yasushi

    1988-01-01

    The LMFBR Super Long-Life Core (SLLC) concept has evolved over the last few years as one of the targets of innovative approaches for future FBR cost reduction. An idea for SLLC has been developed wherein the core lifetime is extended up to the plant life of about 30 years by applying the radially and axially multi-zoned core concept (the improved homogeneous core concept). The main purpose of the present study is placed on the evaluation of neutronic feasibility of the 1000 MWe class SLLC concept. The core size of the present SLLC, which is approximately 3 to 4 times as large as those of the current 1000 MWe core design, was determined by the limit of the maximum fast neutron fluence level, which was tentatively assumed to be 5-6x10 23 nvt as the target of the future development of advanced cladding materials. Emphasis is placed on the discussion of neutronic performances of cores with oxide fuels rather than metal or carbide fuels. The present study has shown that proper zoning of the different plutonium enrichment fuels at the initial core makes it possible to achieve small enough reactivity loss during 30-year burnup while satisfying mild variation of the subassembly power distributions using a higher fuel volume fraction of about 50%. Effects of important neutronic parameters on the core performances are also discussed. (orig.)

  18. Review on Japanese activities in the field of maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Tsuchiya, T.; Fukuda, T.; Sato, M.; Okabayashi, K.; Takahashi, T.

    2002-01-01

    Summary of Japanese activities on maintenance and repair of LMFBR steam generators (SG) is described in this paper. The concept (adoption of helical coil tube etc.) of MONJU SG was established in conceptional design started from 1968, and research and development (R and D) program was prepared. Parallel with basic studies such as material, welding, sodium water reaction and etc., overall verification tests using mock up SGs were conducted. As the first step, 1 Mw SG with two active helical tubes (and eight dummy tubes) was fabricated and operated, and many maintenance and repair experiences were accumulated through two small water leak troubles. Two 50Mw SGs, 1/5 scale of MONJU SG, were constructed and operated for long time. Post test examinations were carried out for No.1 50 Mw SG and feasibility of this type of SG was confirmed. In regard to maintenance and repair techniques, explosive and welding method for tube plugging and UT and ECT techniques for inspection of tube integrity are under development. Overall verification test for on-site and in-factory maintenance and repair techniques was conducted using No.2 50Mw SG evaporator and applicability of those techniques to real plant was evaluated. Many experiences were accumulated for removal and cleaning of sodium water reaction products after sodium water reaction in the cooling system and pressure relief system, using the Large Sodium Water Reaction Test Facility (SWAT-1 and 3). (author)

  19. Ultrasonic inspection for wastage in the LMFBR steam generator due to sodium--water reactions

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, L.

    1977-01-01

    As part of a program to study the results of large sodium-water reactions in the LMFBR Steam Generator, a boreside ultrasonic inspection device was developed to measure the wall thickness and diameter of the 2- 1 / 4 Cr-1 Mo, 0.397 in. I.D. steam tubes. The reaction was created in a near prototype steam generator by guillotine-type rupture of a steam tube, while the generator was at operating conditions. Wastage occurred on the surrounding tubes due to the high temperature reaction. The UT test instrument was designed to operate with a 15 MHz transducer in the pulse-echo shear-wave mode, with a sampling rate of 10 4 /sec. System outputs are diameter, wall thickness, attitude and axial position of the transducer. All are displayed digitally and may be recorded. Measurements are fed into a computer for later retrieval, and/or cascaded outputs into an x-y recorded displaying either out-of-limit or thickness data. The UT data taken in this experiment were consistent with physical measurements on a tube which was removed from the generator after the test. A machined flat 1 / 8 -inch long and 0.002-inch deep could readily be detected

  20. Boreside rotating ultrasonic tester for wastage determination of LMFBR-type steam generator tubes

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, H.L.

    1979-01-01

    Large sodium-water reaction (SWR) leak tests are being run in near-prototypic steam generators at prototypic plant conditions of the Liquid Metal Fast Breeder Reactor (LMFBR). These tests simulate various types of steam tube failure at predetermined locations. A SWR results in a highly energetic-exothermic-caustic reaction which erodes neighboring tubes. A boreside-rotating ultrasonic inspection device was developed to measure wall thickness and inside diameter of the 2/one quarter/Cr-1 Mo, 10.1 mm I.D. steam tubes. Rotation of the UT beam yields a complimentary scan of the full tube in a single pass. The UT system was designed with a 15 MHz transducer in pulse-echo compression-wave mode at a pulse rate of 10,000/second. The UT beam is rotated at 20 r/s on a 1.27 mm pitch. System outputs are diameter, wall thickness, attitude, and axial position. Measurements are processed, then fed to a CRT and computer for later retrieval and plotting

  1. Effects of governing parameters on steady-state inter-wrapper flow in an LMFBR

    International Nuclear Information System (INIS)

    Moriya, Shoichi

    2001-01-01

    Hydraulic experiments were performed using a 1/8th scale rectangular model, based on a Japanese demonstration fast breeder reactor design, in order to study fundamental characteristics of interwrapper flows occurring under steady state conditions in an LMFBR. The steady state interwrapper flow of which direction was downward in the center region and upward in the peripheral region of a core barrel was observed because of the radial static pressure gradient in the upper part of the core barrel, produced by a core blockage effect resulting from an above core structure with a perforated skirt. Thermal stratification phenomena were moreover observed in the interwrapper region, created by the hot steady state interwrapper flow from an upper plenum and the cold leakage flow through the separated plate of the core barrel. The thermal interface was generated in higher part of the core barrel when the core blockage effect was smaller and Richardson number and the leakage flow rate ratio were larger. Significant temperature fluctuations occurred in the peripheral region of the core barrel, when the difference between the interface elevations in the center and peripheral regions of the core barrel was enough large. (author)

  2. Comparative analysis of a hypothetical loss-of-flow accident in an irradiated LMFBR core using different computer models for a common benchmark problem

    International Nuclear Information System (INIS)

    Wider, H.U.; Devos, J.; Nguyen, H.; Goethem, G. Van.; Miles, K.J.; Tentner, A.M.; Pizzica, P.

    1989-01-01

    This report summarizes the results of an international exercise to compare whole-core accident calculations of the initiation phase of an unprotected LOF accident in a large irradiated LMFBR. The results for the accident phase before pin failure are in rather good agreement except for the fuel pin mechanics predictions. There are also some differences in the sodium boiling calculations but the voiding rates which are of key importance are very similar. The post - failure fuel motion and sodium voiding predictions show significant differences. However, the majority of these calculations agree that temporary fuel accumulations occur which increase the power beyond that caused by sodium voiding alone

  3. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  4. Environmental genomics of "Haloquadratum walsbyi" in a saltern crystallizer indicates a large pool of accessory genes in an otherwise coherent species

    NARCIS (Netherlands)

    Legault, Boris A.; Lopez-Lopez, Arantxa; Alba-Casado, Jose Carlos; Doolittle, W. Ford; Bolhuis, Henk; Rodriguez-Valera, Francisco; Papke, R. Thane

    2006-01-01

    Background: Mature saturated brine (crystallizers) communities are largely dominated (> 80% of cells) by the square halophilic archaeon "Haloquadratum walsbyi". The recent cultivation of the strain HBSQ001 and thesequencing of its genome allows comparison with the metagenome of this taxonomically

  5. Swimming pool cleaner poisoning

    Science.gov (United States)

    Swimming pool cleaner poisoning occurs when someone swallows this type of cleaner, touches it, or breathes in ... The harmful substances in swimming pool cleaner are: Bromine ... copper Chlorine Soda ash Sodium bicarbonate Various mild acids

  6. Swimming pool granuloma

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/001357.htm Swimming pool granuloma To use the sharing features on this page, please enable JavaScript. A swimming pool granuloma is a long-term (chronic) skin ...

  7. LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

  8. Pooled analysis of large and long-term safety data from the human papillomavirus-16/18-AS04-adjuvanted vaccine clinical trial programme

    Science.gov (United States)

    Angelo, Maria-Genalin; David, Marie-Pierre; Zima, Julia; Baril, Laurence; Dubin, Gary; Arellano, Felix; Struyf, Frank

    2014-01-01

    Purpose The purpose of this study is to further evaluate the safety of the human papillomavirus (HPV)-16/18-AS04-adjuvanted vaccine (HPV-16/18-vaccine Cervarix®, GlaxoSmithKline, Belgium) through a pooled analysis of data from 42 completed/ongoing clinical studies. Methods Unsolicited adverse events (AEs) were reported for 30 days after each dose. Medically significant conditions, serious AEs (SAEs), potential immune-mediated diseases (pIMDs) and pregnancy outcomes were captured until study completion. Events leading to subject withdrawal were reviewed. Relative risks compared incidences of spontaneous abortion and pIMDs in controlled studies. Results Thirty one thousand one hundred seventy-three adolescent girls/women received HPV-16/18-vaccine alone (HPV group), 2166 received HPV-16/18-vaccine coadministered with another vaccine and 24 241 were controls. Mean follow-up was 39 months (range 0–113.3). Incidences of unsolicited AEs reported within 30 days after any dose were similar between HPV and Control groups (30.8%/29.7%). During the entire study period, reports of medically significant conditions (25.0%/28.3%) and SAEs (7.9%/9.3%) were also similarly distributed between groups. Deaths were rare: HPV (alone/coadministered) n = 25, controls n = 20 (n = 18 in blinded groups). pIMDs within 1 year were reported by 0.2% of HPV-16/18 vaccinees and controls. For each pIMD event category, no increased relative risks were reported for HPV-16/18 vaccinees versus controls. Coadministration did not change the overall safety profile. Pregnancy outcomes and withdrawal rates were similar between groups. Conclusions Analysis of safety data arising from 57 580 subjects and 96 704 HPV-16/18-vaccine doses shows that the incidences and distribution of AEs were similar among HPV-16/18-vaccine recipients and controls. No new safety signals were identified. The data confirm previous findings that HPV-16/18-vaccine has an acceptable benefit-risk profile in adolescent girls and

  9. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  10. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  11. Structural analysis for LMFBR applications[Indian position paper

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-05-01

    Firstly, we discuss the use of elastic analysis for structural design of LMFBR components. The elastic analysis methods have been used for structural design of the Fast Breeder Test Reactor as well as the proposed prototype Test Breeder Reactor. The design of Fast Breeder Test Reactor which is nearing completion is the same as that of Rapsodie. Nevertheless, the design had to he checked against the latest design codes available, namely the ASME Code case 1592. This paper however, is confined to Structural analysis of PFBR components. The problems faced in the design of some of the components, in particular, the inner vessel (plenum separator) are discussed. As far as design codes are concerned, we make use of ASME Code Section III and the Code Case N-47, for high temperature design. The problem faced in the use of these rules are also described along with the description of analysis. Studies in the field of cyclic loading include extension of Bree's breakdown and plastic cycling criteria for ratchet free operation to biaxial stress fields. In other fields, namely, inelastic analysis, piping analysis in the creep regime etc. we are only at a start.

  12. Experience on detection of leakages in LMFBR-steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Smit, C C

    1975-07-01

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks.

  13. Hydrodynamic analysis of the LMFBR prompt burst excursion (PBE) experiment

    International Nuclear Information System (INIS)

    Young, M.F.

    1977-01-01

    A series of in-pile experiments has been conducted at Sandia Laboratories to provide information on pressure levels and conversion of thermal energy into mechanical work in LMFBR cores during hypothetical, superprompt-critical excursions. Pressures generated in these experiments are recorded by a pressure transducer located at the top and bottom of a sodium channel surrounding a single, fresh UO 2 fuel pin. Work energy conversion is measured by a linear motion transducer connected to a piston at the top of the sodium column. Since the pressure transducers are located fairly far from the location of pin failure, it becomes necessary to determine the effect of channel geometry and piston motion on the observed pressure data. A two-dimensional, hydrodynamic analysis of pressure pulse propagation in the fuel pin-coolant channel geometry was therefore performed using the CSQII computer code. The initial series of PBE experiments consists of single, fresh UO 2 pins surrounded by a sodium-filled or dry-coolant channel contained in a closed test capsule. The capsule is subjected to a maximum pulse in the Annular Core Pulse Reactor (ACPR) resulting in an energy deposition of from 2350 to 2900 J/g (14 and 20 percent enriched pins). The pulse width at half maximum (PWHM) is about 5 ms

  14. Evaluation of high-pressure containment buildings for LMFBR's

    International Nuclear Information System (INIS)

    Armstrong, G.R.

    1981-01-01

    A study was conducted on the use of High Pressure LMFBR Containment Buildings for 1000 MW(e) LMFBRs. Two principal aspects were investigated: accident consequence mitigation and cost. Two types of hypothetical accidents were analyzed to establish consequence mitigation: melt-through and energetic expulsion. Three Containment Building (CB) design pressures were investigated: 69 kPa (10 psig), 207 kPa (30 psig), and 414 kPa (60 psig). Four types of design structures were analyzed to establish cost: steel, steel with confinement building, reinforced concrete, and prestressed/post-tensioned concrete. Results show that: it is within reason that a high pressure containment for a 1000 MW(e) reactor can be fabricated that will retain its integrity during postulated severe hypothetical accidents, if available measures are taken to reduce or prevent hydrogen production and the cost differential between basic high (414 kPa) and low (69 kPa) pressure containments is $10 x 10 6 or less

  15. Experience on detection of leakages in LMFBR-steam generators

    International Nuclear Information System (INIS)

    Smit, C.C.

    1975-01-01

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks

  16. Residual stress effects in LMFBR fracture assessment procedures

    International Nuclear Information System (INIS)

    Hooton, D.G.

    1984-01-01

    Two post-yield fracture mechanics methods, which have been developed into fully detailed failure assessment procedures for ferritic structures, have been reviewed from the point of view of the manner in which as-welded residual stress effects are incorporated, and comparisons then made with finite element and theoretical models of centre-cracked plates containing residual/thermal stresses in the form of crack-driving force curves. Applying the procedures to austenitic structures, comparisons are made in terms of failure assessment curves and it is recommended that the preferred method for the prediction of critical crack sizes in LMFBR austenitic structures containing as-welded residual stresses is the CEGB-R6 procedure based on a flow stress defined at 3% strain in the parent plate. When the prediction of failure loads in such structures is required, it is suggested that the CEGB-R6 procedure be used with residual/thermal stresses factored to give a maximum total stress of flow stress magnitude

  17. Multicell slug flow heat transfer analysis of finite LMFBR bundles

    International Nuclear Information System (INIS)

    Yeung, M.K.; Wolf, L.

    1978-12-01

    An analytical two-dimensional, multi-region, multi-cell technique has been developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for 7, 19, and 37-rod bundles of different geometries and power distributions. The validity of the technique has been verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential clad temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect for practical considerations. Moreover, the knowledge of the local temperature field of the rod bundle leads to the determination of the effective mixing lengths L/sub ij/ for adjacent subchannels of various geometries. It was shown that the implementation of the accurately determined L/sub ij/ into COBRA-IIIC calculations has fairly significant effects on intersubchannel mixing. In addition, a scheme has been proposed to couple the 2-D distributed and lumped parameter calculation by COBRA-IIIC such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a 7-rod bundle and the results of calculation were compared to those of three-dimensional analyses and experimental measurements

  18. LMFBR plant design features for sodium spill and fire protection

    International Nuclear Information System (INIS)

    Palm, R.E.

    1982-01-01

    Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper

  19. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  20. Irradiation effects on low-friction coatings for LMFBR applications

    International Nuclear Information System (INIS)

    Ward, A.L.; Johnson, R.N.; Guthrie, G.L.; Aungst, R.C.

    1975-11-01

    A variety of wear-resistant low-friction materials has been irradiated in the EBR-II in order to assess their reponse to LMFBR environments. Pre- and postirradiation testing and examination efforts have concentrated on candidate materials for application to the wear pads on FTR ducts (fuel, control, and reflector assemblies), and a significant result has been qualification of a proprietary detonation-gun-applied chromium carbide coating which employs a Ni Cr binder. Additional materials such as Inconel-718, Haynes-273, aluminides, and various chromium carbide/binder combinations, and other application processes such as plasma-spray, weld-overlays, diffusion bonding and explosive bonding, have also been studied. The most detailed examinations were conducted on selected chromium carbide coatings and included visual inspection, weight and dimensional measurements, metallography, electron microprobe, epoxy-lift-off, and x-ray diffraction analysis. Chromium carbide coatings applied by the detonation-gun process have demonstrated a marked superiority to those applied by plasma-spray techniques

  1. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  2. Finite element elastic-plastic analysis of LMFBR components

    International Nuclear Information System (INIS)

    Levy, A.; Pifko, A.; Armen, H. Jr.

    1978-01-01

    The present effort involves the development of computationally efficient finite element methods for accurately predicting the isothermal elastic-plastic three-dimensional response of thick and thin shell structures subjected to mechanical and thermal loads. This work will be used as the basis for further development of analytical tools to be used to verify the structural integrity of liquid metal fast breeder reactor (LMFBR) components. The methods presented here have been implemented into the three-dimensional solid element module (HEX) of the Grumman PLANS finite element program. These methods include the use of optimal stress points as well as a variable number of stress points within an element. This allows monitoring the stress history at many points within an element and hence provides an accurate representation of the elastic-plastic boundary using a minimum number of degrees of freedom. Also included is an improved thermal stress analysis capability in which the temperature variation and corresponding thermal strain variation are represented by the same functional form as the displacement variation. Various problems are used to demonstrate these improved capabilities. (Auth.)

  3. Expanded bed adsorption as a fast technique for the large-scale purification of the complete isoform pool of Ber e 1, the major allergen from Brazil nuts.

    NARCIS (Netherlands)

    Boxtel, van E.L.; Koningsveld, van G.A.; Koppelman, S.J.; Broek, van den L.A.M.; Voragen, A.G.J.; Gruppen, H.

    2006-01-01

    A new, fast, large-scale purification method for Ber e 1, the major allergen from Brazil nuts, using expanded bed adsorption (EBA) chromatography, is presented. Using EBA, crude extracts can be applied to a fluidized column, which allows the unhindered passage of particulate impurities, thereby

  4. Importance of regional species pools and functional traits in colonization processes: predicting re-colonization after large-scale destruction of ecosystems

    NARCIS (Netherlands)

    Kirmer, A.; Tischew, S.; Ozinga, W.A.; Lampe, von M.; Baasch, A.; Groenendael, van J.M.

    2008-01-01

    Large-scale destruction of ecosystems caused by surface mining provides an opportunity for the study of colonization processes starting with primary succession. Surprisingly, over several decades and without any restoration measures, most of these sites spontaneously developed into valuable biotope

  5. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Roake, W.E.

    1977-01-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  6. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    International Nuclear Information System (INIS)

    Minden, C. v.; Schultheiss, G.F.

    1982-01-01

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

  7. Comparative analysis of LMFBR licensing in the United States and other countries - notably France. Executive summary

    International Nuclear Information System (INIS)

    Golay, M.W.; Castillo, M.

    1981-01-01

    The safety-related design aspects and licensing experiences of LMFBR projects in other democratic countries have been studied and contrasted to those in the United States in order to understand the importance of different approaches to safety, and also to understand better the system of the United States. The regulatory systems and LMFBR programs of France and the United States are contrasted in detail, and that of West Germany is also studied. The programs of Japan and the United Kingdom receive considerably less attention, and that of the Soviet Union is ignored

  8. Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

    International Nuclear Information System (INIS)

    Shin, Y.S.; Wambsganss, M.W.

    1975-05-01

    This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

  9. Cover gas seals. 11 - FFTF-LMFBR seal-test program, January-March 1974

    International Nuclear Information System (INIS)

    Kurzeka, W.; Oliva, R.; Welch, F.

    1974-01-01

    The objectives of this program are to: (1) conduct static and dynamic tests to demonstrate or determine the mechanical performance of full-size (cross section) FFTF fuel transfer machine and reactor vessel head seals intended for use in a sodium vapor - inert gas environment, (2) demonstrate that these FFTF seals or new seal configuration provide acceptable fission product and cover gas retention capabilities at LMFBR Clinch River Plant operating environmental conditions other than radiation, and (3) develop improved seals and seal technology for the LMFBR Clinch River Plant to support the national objective to reduce all atmospheric contaminations to low levels

  10. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

    1977-04-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.

  11. Thermal interactions of a molten core debris pool with surrounding structural materials

    International Nuclear Information System (INIS)

    Baker, L. Jr.; Cheung, F.B.; Farhadieh, R.; Stein, R.P.; Gabor, J.D.; Bingle, J.D.

    1979-01-01

    Analytical and experimental results on individual aspects of the overall problem of the interaction of a large mass of LMFBR core debris with concrete or other materials are reviewed. Results of recent heat transfer experiments with molten UO 2 have indicated the importance of internal thermal radiation and methods to take account of this are developed. Effects of gas release and density difference are considered. The GROWS-2 Code is used to illustrate the effects of various assumptions

  12. Contribution of the CEC in structural analysis applied to LMFBR problems

    International Nuclear Information System (INIS)

    Larsson, L.H.; Terzaghi, A.

    1983-01-01

    This paper presents both the activity of DG XII in field of Codes and Standards (harmonization) and the research activity carried out at the JRC in Ispra. The first part describes the activity performed in the field of structural analysis by the Fast Reactor Coordinating Committee of the CEC and its Working Group Codes and Standards. This activity, which is aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, has made good progress in most areas. Results from recent inelastic and seismic benchmark calculations are presented as well as future computational exercises and investigations related to piping analysis, defect analysis, material behaviour and life prediction at elevated temperature. In the second part of the paper results of recent research and future plans in the area of structural mechanics at the JRC Ispra are discussed. In the past years, a large effort was devoted to the COVA (code validation) program intended to validate dynamic fluid/structure codes necessary for predicting the response of LMFBR containments. The main conclusions that can be drawn from COVA which finishes this year are presented, and some still open questions related to the prediction of containment response to an HCDA are discussed. The paper then describes the identification technique which is applied for the determination of constitutive equations for the dynamic behaviour of materials. In the field of fracture mechanics JRC has mostly concentrated its efforts on the elastic-plastic fracture toughness properties of irradiated austenitic steels. In the future, also dynamic ductile fracture problems will be investigated, for these a large dynamic test facility with a max. force of 5 MN will be used. The numerical analysis methods associated with these tests are discussed. (author)

  13. Solar swimming pool

    Energy Technology Data Exchange (ETDEWEB)

    1985-01-01

    This report examines the feasibility of using solar collectors to heat the water in a previously unheated outdoor swimming pool. The solar system is used in conjunction with a pool blanket, to conserve heat when the pool is not in use. Energy losses through evaporation can be reduced by as much as 70% by a pool blanket. A total of 130 m{sup 2} of highly durable black synthetic collectors were installed on a support structure at a 30{degree} angle from the horizontal, oriented to the south. Circulation of pool water though the collectors, which is controlled by a differential thermostat, was done with the existing pool pump. Before installation the pool temperature averaged 16{degree}C; after installation it ranged from 20{degree} to 26{degree}C. It was hard to distinguish how much pool heating was due to the solar system and how much heat was retained by the pool blanket. However, the pool season was extended by five weeks and attendance tripled. 2 figs.

  14. About application of method of atomic and absorption spectroscopy for investigation of large di cuspid mollusks as test-object of environmental conditions in Zaragshan pool

    International Nuclear Information System (INIS)

    Izzatulaev, Z.I.

    1999-01-01

    The knowledge of water ecosystems evolution laws requires for thorough investigations of chemical and elementary composition of aquatic organisms and the environment, which is of a great importance for the regions affected by technical factors influence. Therefore, we started investigating the large di cuspid molluscs as water bio filters for the first time.The investigations were conducted using different species of mollusks of Unionid and Korbikulid tribes.Contents of lead, zinc, cadmium, copper, magnesium, iron, chrome and mercury in the molluscs were determined using atomic and absorption spectrophotometer. High content of microelements in soft tissues and shell limestone is explained by the fact that they are directly absorbed in the food chain.The investigations data are not final. The investigations will be conducted in the future

  15. Dynamic response of single hexagonal LMFBR core subassembly wrappers

    Energy Technology Data Exchange (ETDEWEB)

    Ash, J. E.; Marciniak, T. J.; (Argonne National Lab., IL (United States))

    1977-07-01

    To analyze the dynamic structural response of the LMFBR core subassembly hexagonal wrappers to postulated local energy releases and the sensitivity of the response to variations in both the pressure loading and the material properties of the stainless steel, a finite-element computer code STRAW has been developed. A series of experiments was performed to study the effects of variations in material properties. The amount of coldworking to which the Type 316 stainless steel is subjected has a strong influence upon the ductility and the elastic yield point. The usual fabrication process produced a nominally 20% coldworking with a yield point of about 680 MPa. By designing a special set of dies for the drawing process, a very low ductility hexcan was produced for which the yield point was raised to 820 MPa. Conversely, the yield point was lowered to 170 MPa by a solution annealing process producing a highly ductile test hexcan. A metallurgical study was conducted to find a representative brittle simulant material for the irradiated end-of-life steel properties. An aging treatment for Type 446 stainless steel was developed which reproduced the expected tensile-flow behavior of the in-pile subassembly. Further study is underway to investigate the fracture properties of the simulant material. The pressure pulses were generated by the controlled expansion of high-pressure detonation poducts from low-density explosives detonated inside a vented steel cannister. The orifice configuration of the cannister and the charge mixture ratio were designed to produce two specified pulse shapes. A charge containing 37,7 g PETN mixed with 35 wt % inert, hollow-glass microballoons developed a pressure pulse peak of 9.5 MPa at 1.0 ms. Increasing the PETN to 41 g resulted in a 14.6 MPa peak pressure, and increasing the explosive concentration to 90 wt % in the mixture increased the burning rate and the pulse risetime, so that the peak occurred at 0.6 ms.

  16. Fuel pin response to an overpower transient in an LMFBR

    International Nuclear Information System (INIS)

    Grosberg, A.J.; Head, J.L.

    1979-01-01

    This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

  17. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  18. Input parameters to codes which analyze LMFBR wire-wrapped bundles

    International Nuclear Information System (INIS)

    Hawley, J.T.; Chan, Y.N.; Todreas, N.E.

    1980-12-01

    This report provides a current summary of recommended values of key input parameters required by ENERGY code analysis of LMFBR wire wrapped bundles. This data is based on the interpretation of experimental results from the MIT and other available laboratory programs

  19. The state of art of the methods for thermohydraulics design of LMFBR fuel elements

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1981-09-01

    The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

  20. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  1. LARA: Expert system for acoustic localization of robot in a LMFBR

    International Nuclear Information System (INIS)

    Lhuillier, C.; Malvache, P.

    1986-12-01

    The expert system LARA (Acoustic Localization of Autonomic Robot) has been developed to show the interest of introducing artificial intelligency for fine automatic positioning of refuelling machine in a LMFBR reactor. LARA which is equipped with an acoustic detector gives rapidly a good positioning on the fuel [fr

  2. LMFBR safety. 6. Review of current issues and bibliography of literature (1977)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1978-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development. Selected bibliographic information on LMFBRs relative to the development and safety of the breeder reactor is presented for the year 1977. The bibliography consists of approximately 198 abstracts covering research and development, operating experience, and design practices. Keyword, author, and permuted-title indexes are included for completeness

  3. Subassembly faults diagnostic of an LMFBR type reactor by the measurement of temperature noise

    International Nuclear Information System (INIS)

    Kokorev, B.V.; Palkin, I.I.; Turchin, N.M.; Pallagi, D.; Horanyi, S.

    1979-09-01

    The subassembly faults detection possibility by temperature noise analysis of an LMFBR is described. The paper contains the results of diagnostical examinations obtained on electrically heated NaK test rigs. On the basis of these results the measurement of temperature noise RMS value seems to be a practicable method to detect local blockages in an early phase. (author)

  4. LMFBR safety program. Annual technical progress report. Government fiscal year, 1977

    International Nuclear Information System (INIS)

    1977-01-01

    Information is presented concerning the development of the SOMIX-1 computer code for sodium drop burning analysis; experimental analysis of burning sodium drops; aerosol leakage from containment buildings; high-temperature-concentration aerosols; aerosol source term from vaporized fuel; properties of high-temperature fuel mixtures; and development of the COMRADEX computer code for analysis of radiological doses in the environment from LMFBR accidents

  5. Review of pertinent thermal-hydraulic data for LMFBR core natural circulation analyses

    International Nuclear Information System (INIS)

    Bishop, A.A.; Coffield, R.D. Jr.; Markley, R.A.

    1980-01-01

    A literature review and summary of significant data is presented relative to LMFBR core natural convection cooling analysis. First, a brief review of computer codes and respective input data needs is made, significant data areas are then addressed and data for verifying the code calculations are described. Recommendations and conclusions with regard to the data are included

  6. A survey of the French creep-fatigue design rules for LMFBR

    International Nuclear Information System (INIS)

    Tribout, J.; Cordier, G.; Moulin, D.

    1987-01-01

    The paper provides a survey of the creep-fatigue design rules for the LMFBR in France. These rules are the ones currently implemented in French component manufacturing. The background of each item is discussed and the trends for improvements currently investigated are described. The creep-fatigue rules apply to elastic analysis only. (orig.)

  7. A miniature inductive temperature sensor to monitor temperature noise in the coolant of an LMFBR

    International Nuclear Information System (INIS)

    Dean, S.A.; Sandham, C.W.

    1980-01-01

    A description is given of the design and performance of miniature inductive sensors developed to monitor fast temperature fluctuations in the sodium coolant above the core of a LMFBR. These instruments, designed to be installed within existing thermocouple containment thimbles, also provide a steady-state temperature indication for reactor control purposes. (author)

  8. Theoretical study and experimental investigation of mixed and natural circulation in LMFBR core subassemblies

    International Nuclear Information System (INIS)

    Leteinturier, D.; Blanc, D.; Menant, B.; Basque, G.

    1980-02-01

    A presentation is made of theoretical and experimental studies carried out in France on mixed and natural convection in LMFBR wire wrapped bundles. Two codes are described, one for mixed convection THERNAT and the other for natural convection BACCHUS. THe related experimental program FETUNA, with electrically heated bundles in sodium loops, is also presented

  9. The water vapor nitrogen process for removing sodium from LMFBR components

    Energy Technology Data Exchange (ETDEWEB)

    Crippen, M D; Funk, C W; Lutton, J M [Hanford Engineering Development Laboratory, Richland (United States)

    1978-08-01

    Application and operation of the Water Vapor-Nitrogen Process for removing sodium from LMFBR components is reviewed. Emphasis is placed on recent efforts to verify the technological bases of the process, to refine the values of process parameters and to ensure the utility of the process for cleaning and requalifying components. (author)

  10. Swimming-pool piles

    International Nuclear Information System (INIS)

    Trioulaire, M.

    1959-01-01

    In France two swimming-pool piles, Melusine and Triton, have just been set in operation. The swimming-pool pile is the ideal research tool for neutron fluxes of the order of 10 13 . This type of pile can be of immediate interest to many research centres, but its cost must be reduced and a break with tradition should be observed in its design. It would be an advantage: - to bury the swimming-pool; - to reject the experimental channel; - to concentrate the cooling circuit in the swimming-pool; - to carry out all manipulations in the water; - to double the core. (author) [fr

  11. The benefits and problems of base seismic isolation for LMFBR reactor plants

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1988-01-01

    The use of seismic isolation as an approach to aseismic design has gained increasing interest as a viable and efficient engineering solution to earthquake ground motion both within and outside of the nuclear field. Seismic isolation design is fundamentally different from conventional design practice. In the conventional approach, seismic loads are resisted by making the structures, equipment, piping, and associated supports strong enough to resist seismic loads and to provide high levels of ductility. The use of seismic isolation approaches the problem by decoupling the structure (and its contents) from the seismic input resulting from ground shaking. Because LMFBR systems operate at virtually atmospheric pressure, vessels, piping, and associated components tend to be quite thin-walled. The problem is that these thin-walled items have little inherent resistance to earthquake effects and are vulnerable to seismic load effects. As a result, earthquake loads have an even greater influence on LMR designs than they already are in LWR plants. The potential benefits of seismic isolation for an LMR plant are considerable, including minimization of high-cost commodities such as stainless steel, large reductions in internal equipment loads, increased margins of safety for beyond-design-basis loads, and enhancement of plant standardization design. There are, of course, a number of issues and concerns in the use of seismic isolation for a nuclear power plant. These issues cover a number of items such as the lack of experience in actual earthquakes, effects of long-period ground motion, effect of vertical loads, traveling waves, and other related concerns. This paper presents an evaluation of the benefits and problems in the use of seismic isolation in LMR plants. 12 refs, 7 figs

  12. Synthesis Report on the understanding of failed LMFBR fuel element performance

    International Nuclear Information System (INIS)

    Plitz, H.; Bagley, K.; Harbourne, B.

    1990-07-01

    In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

  13. Spent fuel storage pool

    International Nuclear Information System (INIS)

    Murakami, Naoshi.

    1996-01-01

    Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)

  14. Collaborative Car Pooling System

    OpenAIRE

    João Ferreira; Paulo Trigo; Porfírio Filipe

    2009-01-01

    This paper describes the architecture for a collaborative Car Pooling System based on a credits mechanism to motivate the cooperation among users. Users can spend the accumulated credits on parking facilities. For this, we propose a business model to support the collaboration between a car pooling system and parking facilities. The Portuguese Lisbon-s Metropolitan area is used as application scenario.

  15. SIMMER-I: an S/sub n/, Implicit, Multifield, Multicomponent, Eulerian, Recriticality code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Bell, C.R.; Bleiweis, P.B.; Boudreau, J.E.; Parker, F.R.; Smith, L.L.

    1976-08-01

    Physical models, numerical methods, and program description are presented for SIMMER-I, a computer program which predicts the neutronic and fluid dynamic behavior of an LMFBR during a hypothetical core disruptive accident

  16. Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

    International Nuclear Information System (INIS)

    Salvatores, M.

    1986-09-01

    Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

  17. 13 CFR 120.611 - Pools backing Pool Certificates.

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Pools backing Pool Certificates. 120.611 Section 120.611 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Secondary Market Certificates § 120.611 Pools backing Pool Certificates. (a) Pool characteristics. As set...

  18. Swimming Pool Safety

    Science.gov (United States)

    ... Spread the Word Shop AAP Find a Pediatrician Safety & Prevention Immunizations All Around At Home At Play ... Español Text Size Email Print Share Swimming Pool Safety Page Content ​What is the best way to ...

  19. Swimming pool special; Zwembadspecial

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-05-15

    This issue includes a few articles and messages on the use of heat pump systems in swimming pools. [Dutch] Dit nummer bevat onder meer een paar artikelen over het gebruik van warmtepompsystemen in zwembaden.

  20. Studies of thermal stratification in water pool

    International Nuclear Information System (INIS)

    Verma, P.K.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.

    2015-01-01

    Large water pools are used as a heat sink for various cooling systems used in industry. In context of advance nuclear reactors like AHWR, it is used as ultimate heat sink for passive systems for decay heat removal and containment cooling. This system incorporates heat exchangers submerged in the large water pool. However, heat transfer by natural convection in pool poses a problem of thermal stratification. Due to thermal stratification hot layers of water accumulate over the relatively cold one. The heat transfer performance of heat exchanger gets deteriorated as a hot fluid envelops it. In the nuclear reactors, the walls of the pool are made of concrete and it may subject to high temperature due to thermal stratification which is not desirable. In this paper, a concept of employing shrouds around the heat source is studied. These shrouds provide a bulk flow in the water pool, thereby facilitating mixing of hot and cold fluid, which eliminate stratification. The concept has been applied to the a scaled model of Gravity Driven Water Pool (GDWP) of AHWR in which Isolation Condensers (IC) tubes are submerged for decay heat removal of AHWR using ICS and thermal stratification phenomenon was predicted with and without shrouds. To demonstrate the adequacy of the effectiveness of shroud arrangement and to validate the simulation methodology of RELAP5/Mod3.2, experiments has been conducted on a scaled model of the pool with and without shroud. (author)

  1. Plant-life extension planning for an operating LMFBR

    International Nuclear Information System (INIS)

    King, R.W.

    1985-01-01

    The study concluded that continued EBR-II operation is certainly feasible for well beyond 10 more years, and that continued demonstration of the unique inherent safety and operability features of a pool-type liquid-metal-cooled reactor and the demonstration of a reasonable operating lifetime are very important and will provide invaluable information for the design and development of the next generation nuclear power plants

  2. Evaluation of air cleaning system concepts for emergency use in LMFBR plants

    International Nuclear Information System (INIS)

    Hilliard, R.K.; McCormack, J.D.; Postma, A.K.

    1976-12-01

    Nineteen different air cleaning concepts are arranged into twenty-four systems and evaluated for use as accident mitigating systems in LMFBR plants. Both single, low-leakage containment plants and once-through operation applicable to containment/confinement plants are considered. Plant characteristics affecting air cleaning requirements are defined for 1000 MW(e) plants and a sodium and radiological release term is postulated. The accident conditions under which the emergency air cleaning system (EACS) must function is established by use of SOFIRE-II and HAA-3B computer codes. Criteria are developed for evaluating the various systems and for assigning comparative ratings. The numerical ratings are combined with information on cost and development potential to arrive at recommendations for the most promising systems. The conclusion is made that reliable and effective systems are feasible for use as engineered safety features for LMFBR plants, but that development effort is required for all the air cleaning concepts evaluated

  3. Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    2002-01-01

    The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

  4. Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

    1979-01-01

    A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

  5. LMFBR operational and experimental local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    1980-01-01

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS- and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  6. LMFBR operational and experimental in-core local-fault experience, primarily with oxide fuel elements

    International Nuclear Information System (INIS)

    Warinner, D.K.

    Case-by-case reviews of selective world experience with severe local faults, particularly fuel failure and fuel degradation, are reviewed for two sodium-cooled thermal reactors, several LMFBRs, and LMFBR-fuels experiments. The review summarizes fuel-failure frequency and illustrates the results of the most damaging LMFBR local-fault experiences of the last 20 years beginning with BR-5 and including DFR, BOR-60, BR2's MFBS-and Mol-loops experiments, Fermi, KNK, Rapsodie, EBR-II, and TREAT-D2. Local-fault accommodation is demonstrated and a need to more thoroughly investigate delayed-neutron and gaseous-fission-product signals is highlighted in view of uranate formation, observed blockages, and slow fuel-element failure-propagation

  7. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle

  8. Acoustics and voiding dynamics during SLSF simulations of LMFBR undercooling transients

    International Nuclear Information System (INIS)

    Anderson, T.T.; Kuzay, T.M.; Marr, W.W.; Miles, K.J.; Pedersen, D.R.; Thompson, D.H.; Wilson, R.E.

    1978-01-01

    The SLSF is the largest U.S. in-reactor test vehicle for steady-state and transient experiments in an environment typical of a LMFBR core. The SLSF experiment program, sponsored by the Department of Energy, contributes to the LMFBR safety assurance program by providing data on key phenomena that occur during postulated reactor accidents. This paper describes completed SLSF experiments, in-core instrumentation used, and methods of data interpretation to determine sodium boiling and voiding dynamics. Boiling inception is shown to be identifiable from several types of in-core instruments. Location of the boiling front and void growth derived from experimental data are compared with analytical predictions. These and other data form the basis to improve understanding of accidents and to validate or guide the development of accident analysis methods

  9. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors.

  10. Report of the IAEA advisory group meeting on LMFBR fuel reprocessing

    International Nuclear Information System (INIS)

    1976-05-01

    A summary of the papers and discussions of the meeting is presented, reviewing the status of development in LMFBR fuel reprocessing and focusing attention on important problem areas. The following topics are discussed: Transport, storage and removal of sodium; decladding and shearing; dissolution; Purex process; fluoride volatility method; off-gas purification; waste disposal. Status reports of national programmes of Belgium, France, Federal Republic of Germany, Italy, Japan, United Kingdom, USSR and USA are included

  11. LMFBR conceptual design study: an overview of environmental and safety concerns

    International Nuclear Information System (INIS)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE

  12. A survey of LMFBR cavitation technology in the U.S.A

    International Nuclear Information System (INIS)

    Cha, Y.S.; Huebotter, P.R.; Hopenfeld, J.

    1976-01-01

    Several experimental programmes of a basic and applied nature were established in the USA in order to develop guidelines to ensure design and operation of LMFBR hydraulic components free from cavitation and/or cavitation damage. As of March 1976, most of these experimental programs are still in progress. Each programme is briefly described. The available interium data are presented. References that are relevant are provided

  13. LMFBR flexible pipe joint development program. Annual technical progress report, government fiscal year 1977

    International Nuclear Information System (INIS)

    1978-01-01

    Currently, the ASME Boiler and Pressure Vessel Code does not allow the use of flexible pipe joints (bellows) in Section III, Class 1 reactor primary piping systems. Studies have shown that the primary piping loops of LMFBR's could be simplified by using these joints. This simplification translates directly into shorter primary piping runs and reduced costs for the primary piping system. Further cost savings result through reduced vault sizes and reduced containment building diameter. In addition, the use of flexible joints localizes the motions from thermally-induced piping growth into components which are specifically designed to accommodate this motion. This reduces the stress levels in the piping system and its components. It is thus economically and structurally important that flexible piping joints be available to the LMFBR designer. The overall objective of the Flexible Joint Program is to provide this availability. This will be accomplished through the development of ASME rules which allow the appropriate use of such joints in Section III, Class 1 piping systems and through the development and demonstration of construction methods which satisfy these rules. The rule development includes analytic and testing methodology formulations which will be supported by subscale bellows testing. The construction development and demonstration encompass the design, fabrication, and in-sodium testing of prototypical LMFBR plant-size flexible pipe joints which meet all ASME rule requirements. The satisfactory completion of these developmental goals will result in an approved flexible pipe joint design for the LMFBR. Progress is summarized in the following efforts undertaken during 1977 to accomplish these goals: (1) code case support, (2) engineering and design, (3) material development, (4) testing, and (5) manufacturing development

  14. LMFBR conceptual design study: an overview of environmental and safety concerns

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D.L.

    1981-06-01

    The US Department of Energy (DOE) initiated the Liquid Metal Fast Breeder (LMFBR) Conceptual Design Study (CDS) with the objective of maintaining a viable breeder option. The project is scheduled to be completed in FY-1981 but decisions regarding plant construction will be delayed until at least 1985. This report provides a review of the potential environmental and safety engineering concerns for the CDS and recommends specific action for the Environmental and Safety Engineering Division of DOE.

  15. Comparative study of heterogeneous and homogeneous LMFBR cores in some accident conditions

    International Nuclear Information System (INIS)

    Renard, A.; Evrard, G.

    1978-01-01

    An heterogeneous design and a homogeneous one of a LMFBR core with the same power and similar dimensions are compared from the safety point-of-view. The comparison is performed for several accident conditions, such as Loss-of-Flow and Transient Overpower, with the same failure criteria and model assumptions for both cores. Qualitative trends are deduced from the behaviour of the core designs in the investigated transient conditions. (author)

  16. Seismic criteria studies and analyses. Quarterly progress report No. 3. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    1975-01-03

    Information is presented concerning the extent to which vibratory motions at the subsurface foundation level might differ from motions at the ground surface and the effects of the various subsurface materials on the overall Clinch River Breeder Reactor site response; seismic analyses of LMFBR type reactors to establish analytical procedures for predicting structure stresses and deformations; and aspects of the current technology regarding the representation of energy losses in nuclear power plants as equivalent viscous damping.

  17. Comprehensive method of common-mode failure analysis for LMFBR safety systems

    International Nuclear Information System (INIS)

    Unione, A.J.; Ritzman, R.L.; Erdmann, R.C.

    1976-01-01

    A technique is demonstrated which allows the systematic treatment of common-mode failures of safety system performance. The technique uses log analysis in the form of fault and success trees to qualitatively assess the sources of common-mode failure and quantitatively estimate the contribution to the overall risk of system failure. The analysis is applied to the secondary control rod system of an early sized LMFBR

  18. Fretting and wear of stainless and ferritic steels in LMFBR steam generators

    International Nuclear Information System (INIS)

    Lewis, M.W.J.; Campbell, C.S.

    1981-01-01

    Steam generators for LMFBR's may be subject to both fretting wear as a result of flow-induced vibrations and to wear from larger amplitude sliding movements from thermal changes. Results of tests simulating the latter are given for stainless and ferritic steels. For the assessment of fretting wear damage, vibration assessments must be combined with data on specific wear rates. Test mechanisms used to study fretting in sodium covering impact, impact-slide and pure rubbing are described and results presented. (author)

  19. Acoustic detection for water/steam leak from a tube of LMFBR steam generator

    International Nuclear Information System (INIS)

    Sonoda, Masataka; Shindo, Yoshihisa

    1989-01-01

    Acoustic leak detector is useful for detecting more quickly intermediate leak than the existing hydrogen detector and is available for identification of leak location on the accident of water/steam leak from a tube of LMFBR steam generator. This paper presents the overview of HALD (High frequency Acoustics Leak Detection) system, which is more sensitive for leak detection and lower cost of equipment for identification of leak location than a low frequency type detector. (author)

  20. Development of an 85,000 gpm (19,303 m3/h) LMFBR primary pump

    International Nuclear Information System (INIS)

    Zerinvary, M.C.; Wagner, E.W.

    1984-01-01

    The development of an 85,000 gpm two-stage primary pump for liquid metal fast breeder reactor (LMFBR) applications is described. The design was supported by air and cavitation model testing of the hyraulics, and development and feature testing of the level control system and the adjustable frequency solid state power supply. Important fabrication and water test items are also discussed, along with some unique assembly tooling requirements

  1. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors

  2. Damping of the radial impulsive motion of LMFBR core components separated by fluid squeeze films

    International Nuclear Information System (INIS)

    Liebe, R.; Zehlein, H.

    1977-01-01

    The core deformation of a liquid metal cooled fast breeder reactor (LMFBR) due to local pressure propagation from rapid energy releases is a complex three-dimensional fluid-structure-interaction problem. High pressure transients of short duration cause structural deformation of the closely spaced fuel elements, which are surrounded by the flowing coolant. Corresponding relative displacements give rise to squeezing fluid motion in the thin layers between the subassemblies. Therefore significant backpressures are produced and the resulting time and space dependent fluid forces are acting on the structure as additional non-conservative external loads. Realistic LMFBR safety analysis of several clustered fuel elements have to account for such flow induced forces. Several idealized models have been proposed to study some aspects of the complex problem. As part of the core mechanics activities at GfK Karlsruhe this paper describes two fluid flow models (model A, model B), which are shown to be suitable for physically coupled fluid-structure analyses. Important assumptions are discussed in both cases and basic equations are derived for one- and two-dimensional incompressible flow fields. The interface of corresponing computer codes FLUF (model A) and FLOWAX (model B) with structural dynamics programs is outlined. Finally fluid-structure interaction problems relevant to LMFBR design are analyzed; parametric studies indicate a significant cushioning effect, energy dissipation and a strongly nonlinear as well as timedependent damping of the structural response. (Auth.)

  3. Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

    1976-01-01

    A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

  4. Effect of reactor size on the breeding economics of LMFBR blankets

    International Nuclear Information System (INIS)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's

  5. A preliminary design study of a pool-type FBR 'ARES' eliminating intermediate heat transport systems

    International Nuclear Information System (INIS)

    Ueda, N.; Nishi, Y.; Kinoshita, I.; Yoshida, K.

    2001-01-01

    An innovative reactor concept 'ARES' (Advanced Reactor Eliminating Secondary system) is proposed to aim at reducing the construction cost of a liquid metal cooled fast breeder reactor (LMFBR). This concept is developed to show the ultimate cost down potential of LMFBR's at their commercial stage. The electrical output is 1500 MW, while the thermal output is 3900 MW. Main components of the primary cooling system are four electromagnetic pumps (EMP) and eight double-wall-tube steam generators (SG). Both of them are installed in a reactor vessel like pool type LMFBR's. An intermediate heat transport system which a previous LMFBR has it eliminated, main components of which are intermediate heat exchangers (IHX), secondary pumps and secondary piping. Further, a high reliable SG could decrease the occurrence of water leak accidents and reduce the related mitigation systems. In this study, structure concept, approach to embody a high reliable SG and accidents analyses are carried out. Flow path configuration is mainly discussed in investigation of the structure concept. In case of a water leak accident in a SG, the fault SG must be isolated to prevent a reaction production from flowing into the core. The measure to cut both inlet and outlet coolant flow paths by siphon-break mechanism is adopted to be consistent with the decay heat removal operation. The safety design approach of the double-wall-tube SG is investigated to limit the accident occurrence below 10 -7 (1/ry). A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to prevent adhesion of the double-wall-tube effectively. The reliability of the tube-to-tube-sheet was evaluated as 10 -10 (1/hr) for an inner tube and 10 -9 (1/hr) for an outer tube with reference to the failure experience of previous SG's. The failure must be detected within 60 to 120 minutes. Finally, a seamless U tube type of double-wall-tube SG is adopted. Transient events due to

  6. Spent fuel pool cleanup and stabilization

    International Nuclear Information System (INIS)

    Miller, R.L.

    1987-06-01

    Each of the plutonium production reactors at Hanford had a large water-filled spent fuel pool to provide interim storage of irradiated fuel while awaiting shipment to the separation facilities. After cessation of reactor operations the fuel was removed from the pools and the water levels were drawn down to a 5- to 10-foot depth. The pools were maintained with the water to provide shielding and radiological control. What appeared to be a straightforward project to process the water, remove the sediments from the basin, and stabilize the contamination on the floors and walls became a very complex and time consuming operation. The sediment characteristics varied from pool to pool, the ion exchange system required modification, areas of hard-pack sediments were discovered on the floors, special arrangements to handle and package high dose rate items for shipment were required, and contract problems ensued with the subcontractor. The original schedule to complete the project from preliminary engineering to final stabilization of the pools was 15 months. The actual time required was about 25 months. The original cost estimate to perform the work was $2,651,000. The actual cost of the project was $5,120,000, which included $150,000 for payment of claims to the subcontractor. This paper summarizes the experiences associated with the cleanup and radiological stabilization of the 100-B, -C, -D, and -DR spent fuel pools, and discusses a number of lessons learned items

  7. Pool water cleaning facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro [Hitachi Ltd., Tokyo (Japan); Asano, Takashi

    1998-05-29

    Only one system comprising a suppression poor water cleaning system (SPCU) and a filtration desalting tower (F/D) is connected for a plurality of nuclear power plants. Pipelines/valves for connecting the one system of the SPCU pump, the F/D and the plurality of nuclear power plants are disposed, and the system is used in common with the plurality of nuclear power plants. Pipelines/valves for connecting a pipeline for passing SP water to the commonly used SPCU pump and a skimmer surge tank are disposed, and fuel pool water is cooled and cleaned by the commonly used SPCU pump and the commonly used F/D. The number of SPCU pumps and the F/D facilities can be reduced, and a fuel pool water cooling operation mode and a fuel pool water cleaning operation mode which were conducted by an FPC pump so far are conducted by the SPCU pump. (N.H.)

  8. Mathematical modelling and simulation of the thermal performance of a solar heated indoor swimming pool

    OpenAIRE

    Mančić Marko V.; Živković Dragoljub S.; Milosavljević Peđa M.; Todorović Milena N.

    2014-01-01

    Buildings with indoor swimming pools have a large energy footprint. The source of major energy loss is the swimming pool hall where air humidity is increased by evaporation from the pool water surface. This increases energy consumption for heating and ventilation of the pool hall, fresh water supply loss and heat demand for pool water heating. In this paper, a mathematical model of the swimming pool was made to assess energy demands of an indoor swimming po...

  9. Main aspects of the design of a support structure of a LMFBR with particular reference to the explosive accident consequences

    International Nuclear Information System (INIS)

    Giuliano, V.; Lazzeri, L.

    1977-01-01

    The aim of this paper is a review of the main aspects of the design of a support structure of a LMFBR tank, with particular reference to the analysis of the non-linear dynamic behaviour of the structure in the plastic range under the effect of an explosive accident within the tank. The structure is composed by a L-shaped flange, which supports the tank, connected by means of nine square beams to a rigid box-type ring, fixed to the concrete. The plug of the tank is connected to the L-shaped figure by means of a group of SS bars. The non-linear dynamic analysis of the explosive accident has been carried out on a lumped mass model, with elastic-plastic elements which simulate main components of the support structure and tank. The impulsive load connected to the explosive accident has been modelled (on the basis of extensive comparative studies carried out) as two triangular pressure impulses acting on the plug and on the botton of the tank. A large amount of results, which describe displacements, velocities and accelerations of the plug, of the tank, and of the support structure, together with the forces and stresses acting on the main structural components are presented and discussed, with particular reference to the influence of the various parameters involved in the analysis. (Auth.)

  10. Distributed Technologies in a Data Pool

    Science.gov (United States)

    Keiser, K.; Conover, H.; Graves, S.; He, Y.; Regner, K.; Smith, M.

    2004-12-01

    A Data Pool is an on-line repository providing interactive and programmatic access to data products through a variety of services. The University of Alabama in Huntsville has developed and deployed such a Data Pool in conjunction with the DISCOVER project, a collaboration with NASA and Remote Sensing Systems. DISCOVER provides long-term ocean and climate data from a variety of passive microwave satellite instruments, including such products as sea-surface temperature and wind, air temperature, atmospheric water vapor, cloud water and rain rate. The Data Pool provides multiple methods to access and visualize these products, including conventional HTTP and FTP access, as well as data services that provide for enhanced usability and interoperability, such as GridFTP, OPeNDAP, OpenGIS-compliant web mapping and coverage services, and custom subsetting and packaging services. This paper will focus on the distributed service technologies used in the Data Pool, which spans heterogeneous machines at multiple locations. For example, in order to provide seamless access to data at multiple sites, the Data Pool provides catalog services for all data products at the various data server locations. Under development is an automated metadata generation tool that crawls the online data repositories regularly to dynamically update the Data Pool catalog with information about newly generated data files. For efficient handling of data orders across distributed repositories, the Data Pool also implements distributed data processing services on the file servers where the data resides. Ontologies are planned to support automated service chaining for custom user requests. The UAH Data Pool is based on a configurable technology framework that integrates distributed data services with a web interface and a set of centralized database services for catalogs and order tracking. While this instantiation of the Data Pool was implemented to meet the needs of the DISCOVER project, the framework was

  11. Liquid sodium pool fires

    Energy Technology Data Exchange (ETDEWEB)

    Casselman, C [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

    1979-03-01

    Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

  12. Income pooling within families

    DEFF Research Database (Denmark)

    Bonke, Jens; Uldall-Poulsen, Hans

    This paper analyses the phenomenon of income-pooling by applying the Danish household expenditure survey, merged with authoritative register information. Responses to additional questions on income sharing among 1696 couples also allows us to analyses whether the intra-household distribution...

  13. Liquid sodium pool fires

    International Nuclear Information System (INIS)

    Casselman, C.

    1979-01-01

    Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

  14. Prevalence, incidence and mortality from cardiovascular disease in patients with pooled and specific severe mental illness: a large-scale meta-analysis of 3,211,768 patients and 113,383,368 controls.

    Science.gov (United States)

    Correll, Christoph U; Solmi, Marco; Veronese, Nicola; Bortolato, Beatrice; Rosson, Stella; Santonastaso, Paolo; Thapa-Chhetri, Nita; Fornaro, Michele; Gallicchio, Davide; Collantoni, Enrico; Pigato, Giorgio; Favaro, Angela; Monaco, Francesco; Kohler, Cristiano; Vancampfort, Davy; Ward, Philip B; Gaughran, Fiona; Carvalho, André F; Stubbs, Brendon

    2017-06-01

    People with severe mental illness (SMI) - schizophrenia, bipolar disorder and major depressive disorder - appear at risk for cardiovascular disease (CVD), but a comprehensive meta-analysis is lacking. We conducted a large-scale meta-analysis assessing the prevalence and incidence of CVD; coronary heart disease; stroke, transient ischemic attack or cerebrovascular disease; congestive heart failure; peripheral vascular disease; and CVD-related death in SMI patients (N=3,211,768) versus controls (N=113,383,368) (92 studies). The pooled CVD prevalence in SMI patients (mean age 50 years) was 9.9% (95% CI: 7.4-13.3). Adjusting for a median of seven confounders, patients had significantly higher odds of CVD versus controls in cross-sectional studies (odds ratio, OR=1.53, 95% CI: 1.27-1.83; 11 studies), and higher odds of coronary heart disease (OR=1.51, 95% CI: 1.47-1.55) and cerebrovascular disease (OR=1.42, 95% CI: 1.21-1.66). People with major depressive disorder were at increased risk for coronary heart disease, while those with schizophrenia were at increased risk for coronary heart disease, cerebrovascular disease and congestive heart failure. Cumulative CVD incidence in SMI patients was 3.6% (95% CI: 2.7-5.3) during a median follow-up of 8.4 years (range 1.8-30.0). Adjusting for a median of six confounders, SMI patients had significantly higher CVD incidence than controls in longitudinal studies (hazard ratio, HR=1.78, 95% CI: 1.60-1.98; 31 studies). The incidence was also higher for coronary heart disease (HR=1.54, 95% CI: 1.30-1.82), cerebrovascular disease (HR=1.64, 95% CI: 1.26-2.14), congestive heart failure (HR=2.10, 95% CI: 1.64-2.70), and CVD-related death (HR=1.85, 95% CI: 1.53-2.24). People with major depressive disorder, bipolar disorder and schizophrenia were all at increased risk of CVD-related death versus controls. CVD incidence increased with antipsychotic use (p=0.008), higher body mass index (p=0.008) and higher baseline CVD prevalence (p=0.03) in

  15. Establishment and validation of the model of molten pool in fast reactor

    International Nuclear Information System (INIS)

    Zhou Shufeng; Luo Rui; Wang Zhou; Shi Xiaobo; Yang Xianyong

    2007-01-01

    Running under the beyond design base accidental condition, sodium boiling and dry-out will soon be brought about in LMFBR. If not stopped timely, the fuel pins of the subassembly will be melt and broken to form a molten pool at the bottom of the subassembly. to present a reasonable analysis about the molten pool accident, a method of establishing model according to the mechanism is selected, by which an integral model of the molten pool is established. Validated on the three power groups of BF1 experiments which belong to the France SCARABEE series experimenters, the model shows good results. After compared with the models of GEYSER and BF2 experiments which had been validated before, some conclusions about mechanism of molten pool are derived. Moreover, through comparing the relative parameters such as the discharged heat and the increment of temperature etc., a reasonable analysis about the type of heat transfer is present, on the basis of which some conclusions are derived as well. (authors)

  16. Evolution of the roof insulation conception in French LMFBR

    International Nuclear Information System (INIS)

    Frachet, S.; Pradel, P.

    1986-01-01

    France has built two power producing fast breeders: the 250 MWe PHENIX and the 1200 MWe SUPER-PHENIX-1 (with European participation) and is presently working on the preliminary design of a third 1500 MWe reactor SUPER-PHENIX-2. The upper closures of these pool type reactors are essential structural elements since they fulfill several vital functions. The following points are briefly discussed in this paper: The different facilities which have contributed to validate the SPX1 and SPX2 upper-shuttings: Gulliver, Coca, Germinal; the numerical methods which were perfected to validate the calculation of the penetrations; a heat balance which is used to estimate the temperature of the cover gas; the solution adopted to avoid the aerosols deposits

  17. Application of ultrasonic thermometry in LMFBR safety research

    International Nuclear Information System (INIS)

    Carlson, G.A.; Sullivan, W.H.; Plein, H.G.

    1977-01-01

    Ultrasonic thermometry has many potential applications in reactor safety research, where extremely high temperatures and lack of visual access may preclude the use of conventional diagnostics. An application (the in-core molten fuel pool experiment) will be described in which thoriated tungsten ultrasonic thermometers were used to measure temperatures in UO 2 to incipient melt (2860 0 ). Each thermometer included five sensor elements 10 mm long, providing five temperatures within the UO 2 at various axial locations. The 10 mm spatial resolution is about five times better than previous applications of the technique. Temperature resolution of +-10 0 C was indicated by calibration data. Besides providing temperature data approximately 1000 0 C higher than were obtained with thermocouples, the thermometer yielded valuable axial temperature profile data. Details of the sensors, exciting coils, and signal conditioning electronics will be given

  18. Design-related inherent safety characteristics in large LMFBR power plants

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Barthold, W.P.; Bowers, C.H.; Ferguson, D.R.; Prohammer, F.G.; van Erp, J.B.

    1976-01-01

    Design-related safety-enhancing features such as (1) extended pump coastdown, (2) increased negative reactivity feedbacks, (3) reduced sodium void reactivity, and (4) self-actuated shutdown systems are evaluated. Primary emphasis is placed on preventing or limiting core damage. Attention is also given to features aimed at mitigation of the energetics potential of hypothetical core-disruptive accidents

  19. Pool gateway seal

    International Nuclear Information System (INIS)

    Starr, J.A.; Steinert, L.A.

    1983-01-01

    A device for sealing a gateway between interconnectable pools in a nuclear facility comprising a frame supporting a liquid impermeable sheet positioned in a u-shaped gateway between the pools. An inflatable tube carried in a channel in the periphery of the frame and adjoining the gateway provides a seal therebetween when inflated. A restraining arrangement on the bottom edge of the frame is releasably engagable with an adjacent portion of the gateway to restrict the movement of the frame in the u-shaped gateway upon inflation of the tube, thereby enhancing the seal. The impermeable sheet is formed of an elastomer and thus is conformable to a liquid permeable supportive wall upon application of liquid pressure to the side of the sheet opposite the wall

  20. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Roebert, G.A.

    1978-01-01

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  1. The RCC-MR design code for LMFBR components. A useful basic for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1985-11-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials (Stainless steels), temperature service level (550-600 0 C), loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain

  2. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

  3. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  4. CERN Electronics Pool presentations

    CERN Multimedia

    2011-01-01

    The CERN Electronics Pool has organised a series of presentations in collaboration with oscilloscope manufacturers. The last one will take place according to the schedule below.   Time will be available at the end of the presentation to discuss your personal needs. The Agilent presentation had to be postponed and will be organised later. -     Lecroy: Thursday, 24 November 2011, in 530-R-030, 14:00 to 16:30.

  5. Welding pool measurement using thermal array sensor

    Science.gov (United States)

    Cho, Chia-Hung; Hsieh, Yi-Chen; Chen, Hsin-Yi

    2015-08-01

    Selective laser melting (SLM) is an additive manufacturing (AM) technology that uses a high-power laser beam to melt metal powder in chamber of inert gas. The process starts by slicing the 3D CAD data as a digital information source into layers to create a 2D image of each layer. Melting pool was formed by using laser irradiation on metal powders which then solidified to consolidated structure. In a selective laser melting process, the variation of melt pool affects the yield of a printed three-dimensional product. For three dimensional parts, the border conditions of the conductive heat transport have a very large influence on the melt pool dimensions. Therefore, melting pool is an important behavior that affects the final quality of the 3D object. To meet the temperature and geometry of the melting pool for monitoring in additive manufacturing technology. In this paper, we proposed the temperature sensing system which is composed of infrared photodiode, high speed camera, band-pass filter, dichroic beam splitter and focus lens. Since the infrared photodiode and high speed camera look at the process through the 2D galvanometer scanner and f-theta lens, the temperature sensing system can be used to observe the melting pool at any time, regardless of the movement of the laser spot. In order to obtain a wide temperature detecting range, 500 °C to 2500 °C, the radiation from the melting pool to be measured is filtered into a plurality of radiation portions, and since the intensity ratio distribution of the radiation portions is calculated by using black-body radiation. The experimental result shows that the system is suitable for melting pool to measure temperature.

  6. Intra-patient variability of FDG standardized uptake values in mediastinal blood pool, liver, and myocardium during R-CHOP chemotherapy in patients with diffuse large B- cell lymphoma

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soo Jeong; Yi, Hyun Kyung; Lim, Chae Hong; Cho, Young Seok; Choi, Joon Young; Choe, Yeam Seong; Lee, Kyung Han; Moon, Seung Hwan [Dept. of Nuclear Medicine, Samsung Medical Center, Sungkyunkwan University School of Medicine, Seoul (Korea, Republic of)

    2016-12-15

    {sup 18}F-fluorodeoxyglucose (FDG) PET/CT is useful for staging and evaluating treatment response in patients with diffuse large B-cell lymphoma (DLBCL). A five-point scale model using the mediastinal blood pool (MBP) and liver as references is a recommended method for interpreting treatment response. We evaluated the variability in standardized uptake values (SUVs) of the MBP, liver, and myocardium during chemotherapy in patients with DLBCL. We analyzed 60 patients with DLBCL who received rituximab, cyclophosphamide, doxorubicin, vincristine, and prednisolone (R-CHOP) treatment and underwent baseline, interim, and final FDG PET/CT scans. The FDG uptakes of lymphoma lesions, MBP, liver, and myocardium were assessed, and changes in the MBP and liver SUV and possible associated factors were evaluated. The SUV of the liver did not change significantly during the chemotherapy. However, the SUV{sub mean} of MBP showed a significant change though the difference was small (p = 0.019). SUV{sub mean} of MBP and liver at baseline and interim scans was significantly lower in patients with advanced Ann Arbor stage on diagnosis. The SUV{sub mean} of the MBP and liver was negatively correlated with the volumetric index of lymphoma lesions in baseline scans (r = -0.547, p < 0.001; r = -0.502, p < 0.001). Positive myocardial FDG uptake was more frequently observed in interim and final scans than in the baseline scan, but there was no significant association between the MBP and liver uptake and myocardial uptake. The SUV of the liver was not significantly changed during R-CHOP chemotherapy in patients with DLBCL, whereas the MBP SUV of the interim scan decreased slightly. However, the SUV of the reference organs may be affected by tumor burden, and this should be considered when assessing follow-up scans. Although myocardial FDG uptake was more frequently observed after R-CHOP chemotherapy, it did not affect the SUV of the MBP and liver.

  7. Swimming Pools and Molluscum Contagiosum

    Science.gov (United States)

    ... Travelers’ Health: Smallpox & Other Orthopoxvirus-Associated Infections Poxvirus Swimming Pools Recommend on Facebook Tweet Share Compartir The ... often ask if molluscum virus can spread in swimming pools. There is also concern that it can ...

  8. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-02-24

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included.

  9. PNC status report on leak detector development for LMFBR steam generators

    International Nuclear Information System (INIS)

    Kuroha, M.; Sato, M.

    1984-01-01

    Chemical and acoustic type leak detectors have been developed for detecting a small sodium-water reaction in an LMFBR steam generator. This paper presents a summary of the development. (1) Test results on PNC type in-sodium hydrogen meters including a description of the structure, the long-term reliability and the durability, and the improved meter with an orifice, (2) Development of in-cover gas hydrogen meters, (3) Hydrogen detection tests and analyses, (4) Operating experiences of electrochemical in-sodium oxygen meters, and (5) Basic studies on acoustic characteristics of the sodium-water reaction. (author)

  10. Study of thermal-hydraulic characteristics in an LMFBR intermediate plenum

    International Nuclear Information System (INIS)

    Uotani, M.; Naohara, N.; Kinoshita, I.

    1985-01-01

    Experimental studies using water and liquid metal were conducted in order to investigate the thermal-hydraulic characteristics of an LMFBR intermediate plenum. The present study is an attempt to evaluate the effect of natural convection on the temperature field and to validate the prediction method of temperature profile in a thermally stratified cavity. The experimental results indicated that the effect of the natural convection on flow velocity and heat transfer in the cavity is reduced with increasing the modified stratification parameter. The calculation by FEM code and a simple 1-D model are effective to predict the temperature profile in the cavity

  11. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  12. Analysis of hypothetical LMFBR whole-core accidents in the USA

    International Nuclear Information System (INIS)

    Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

    1978-01-01

    The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

  13. LMFBR safety. 4. Review of current issues and bibliography of literature (1974--1975)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-03-21

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  14. Transference of advanced LMFBR control technology to the aerospace power system program

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Much recent R and D has been devoted to the safety of liquid metal fast breeder reactors (LMFBR's). Part of the resulting technology, especially advanced control systems, appears to be directly transferable to the space nuclear power program. Some of the ideas described herein have been already culminated in successful products that are available for application, e.g. analytical redundancy and fault-tolerant computers. Others, in various stages of R and D, are being developed as elements to support the design goals outlined in the following section, e.g. automated software verification, automated hardware verification, and system validation

  15. Whole-core thermal-hydraulic transient code development and verification for LMFBR analysis

    International Nuclear Information System (INIS)

    Spencer, D.R.

    1979-04-01

    Predicted performance during both steady state and transient reactor operation determines the steady state operating limits on LMFBRs. Unnecessary conservatism in performance predictions will not contribute to safety, but will restrict the reactor to more conservative, less economical steady state operation. The most general method for reducing analytical conservatism in LMFBR's without compromising safety is to develop, validate and apply more sophisticated computer models to the limiting performance analyses. The purpose of the on-going Natural Circulation Verification Program (NCVP) is to develop and validate computer codes to analyze natural circulation transients in LMFBRs, and thus, replace unnecessary analytical conservatism with demonstrated calculational capability

  16. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  17. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  18. LMFBR safety. 3. Review of current issues and bibliography of literature (1972--1974)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1972 through 1974. The bibliography consists of approximately 1380 abstracts covering research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included

  19. Sodium fire studies in France safety tests and applications on an LMFBR

    International Nuclear Information System (INIS)

    Fruchard, Y.; Colome, J.; Malet, J.C.; Berlin, M.; de Cuy, G.D.; Justin, J.; Duco, J.; Fourest, B.

    1976-01-01

    The risk of sodium fires in an LMFBR requires thorough analysis, and the possible consequences of an accidental fire must be accurately determined. Not only must means of prevention and detection be perfected, but techniques must be developed to limit the damage caused by a fire: extinguishment, aerosol containment, protection of reactor structures. The program currently undertaken by the CEA's Nuclear Safety Department covering these problems is described. The major results obtained as well as their application to the SUPER-PHENIX reactor are included

  20. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-01-01

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness

  1. Safety research needs for carbide and nitride fueled LMFBR's. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1975-01-01

    The results of a study initiated at UCLA during the academic year 1974--1975 to evaluate and review the potential safety related research needs for carbide and nitride fueled LMFBR's are presented. The tasks included the following: (1) Review Core and primary system designs for any significant differences from oxide fueled reactors, (2) Review carbide (and nitride) fuel element irradiation behavior, (3) Review reactor behavior in postulated accidents, (4) Examine analytical methods of accident analysis to identify major gaps in models and data, and (5) Examine post accident heat removal. (TSS)

  2. Deposition of inhaled LMFBR-fuel-sodium aerosols in beagle dogs

    International Nuclear Information System (INIS)

    Hackett, P.L.; Mahlum, D.D.; Briant, J.K.; Catt, D.L.; Peters, L.R.; Clary, A.J.

    1980-01-01

    Initial alveolar deposition of LMFBR-fuel aerosols in beagle dogs amounted to 30% of the inhaled activity, but only 5% of the total inhaled activity was deposited in dogs exposed to sodium-fuel aerosols. Aerosol deposition in the gastrointestinal tract amounted to 4% of the initial body burden of fuel-aerosol exposed dogs and 24% of the burden of animals receiving sodium-fuel aerosols. Preliminary analytical data for the dog exposures appear to agree with rodent data for deposition and distribution patterns of aerosols of similar sodium: fuel ratios

  3. LMFBR safety. 2. Review of current issues and bibliography of literature, 1970--1972

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-11-22

    This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1970 through 1972. The bibliography consists of approximately 1620 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  4. Comparative analysis of quality assurance requirements for selected LMFBR components of classes 1, 2 and 3

    International Nuclear Information System (INIS)

    Kleinert, K.P.

    1992-01-01

    The study analyses and compares German, French, British and Italian practices and procedures applied for various LMFBR projects both related to the quality assurance system and related to the particular type of class of component:Class 1: primary reactor vessel; Class 2: Secondary sodium pump; Class 3: Primary cold trap. Various areas of analysis and comparison were selected to identify the underlying concepts of grading of requirements and measures, to identify the similarities and differences, and to give recommendations for further actions concerning quality assurance requirements 60 refs., 21 tabs., 6 figs

  5. Influence of leakage flow on the behaviour of gas behind a blockage in LMFBR subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1980-07-01

    Observations were made of the behaviour of gas behind a uniform porous 21% corner blockage within a pin-bundle of LMFBR subassembly geometry. The main parameter of the experiment was the leakage flow rate through the blockage. The behaviour of gas is significantly influenced by the leakage flow rate. The measured size and residence time of a gas cavity formed behind the blockage are shown and the mechanisms of the gas cavity dispersion by the leakage flow discussed by using a simple model of the liquid flow distribution behind the blockage. (orig.) [de

  6. LMFBR safety. 5. Review of current issues and bibliography of literature (1975--1976)

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1977-06-08

    The current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA), are discussed. Bibliographic information on worldwide LMFBRs relative to the development and safety of the breeder reactor is presented for the period 1975 through 1976. The bibliography consists of approximately 1618 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Keyword, author, and permuted-title indexes are included for completeness.

  7. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    Energy Technology Data Exchange (ETDEWEB)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-08-16

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

  8. LMFBR safety. 1. Review of current issues and bibliography of literature, 1960--1969

    International Nuclear Information System (INIS)

    Buchanan, J.R.; Keilholtz, G.W.

    1976-01-01

    This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness

  9. Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

    International Nuclear Information System (INIS)

    Henslee, S.P.

    1987-03-01

    The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs

  10. Techniques for the thermal/hydraulic analysis of LMFBR check valves

    International Nuclear Information System (INIS)

    Cho, S.M.; Kane, R.S.

    1979-01-01

    A thermal/hydraulic analysis of the check valves in liquid sodium service for LMFBR plants is required to provide temperature data for thermal stress analysis of the valves for specified transient conditions. Because of the complex three-dimensional flow pattern within the valve, the heat transfer analysis techniques for less complicated shapes could not be used. This paper discusses the thermal analysis techniques used to assure that the valve stress analysis is conservative. These techniques include a method for evaluating the recirculating flow patterns and for selecting appropriately conservative heat transfer correlations in various regions of the valve

  11. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  12. Estimated environmental radionuclide transfer and deposition into outdoor swimming pools

    International Nuclear Information System (INIS)

    Tagami, Kazumi; Nagata, Izumi; Sueki, Keisuke

    2014-01-01

    In 2011, a large radioactive discharge occurred at the Fukushima Daiichi nuclear power plant. This plant is located within a climatically temperate region where outdoor swimming pools are popular. Although it is relatively easy to decontaminate pools by refilling them with fresh water, it is difficult to maintain safe conditions given highly contaminated diurnal dust falls from the surrounding contaminated ground. Our objectives in this paper were to conduct daily radioactivity measurements, to determine the quantity of radioactive contaminants from the surrounding environment that invade outdoor pools, and to investigate the efficacy of traditional pool cleaners in removing radioactive contaminants. The depositions in the paper filterable particulates ranged from 0 to 62,5 Bq/m 2 /day, with the highest levels found in the southern Tohoku District containing Fukushima Prefecture and in the Kanto District containing Tokyo Metro. They were approximately correlated with the ground contamination. Traditional pool cleaners eliminated 99% of contaminants at the bottom of the pool, reducing the concentration to 41 Bq/m 2 after cleaning. Authors recommended the deposition or the blown radionuclides into outdoor swimming pools must be considered into pool regulations when the environments exactly polluted with radionuclides. - Highlights: • Deposition into outdoor swimming pool in a habitable areas estimated 72 Bq/m 2 /day. • More than 500 Bq/m 2 /day deposition will exceed our national guideline (10 Bq/l) of swimming pool. • Vacuum pool cleaner eliminates 99% radionuclides deposition

  13. DNA pooling strategies for categorical (ordinal) traits

    Science.gov (United States)

    Despite reduced genotyping costs in recent years, obtaining genotypes for all individuals in a population may still not be feasible when sample size is large. DNA pooling provides a useful alternative to determining genotype effects. Clustering algorithms allow for grouping of individuals (observati...

  14. A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

    International Nuclear Information System (INIS)

    Yoshihisa Shindo; Kazuo Haga

    2005-01-01

    Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

  15. Simulation of LMFBR excursion models by means of ICECO

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Wang, C.Y.; Fistedis, S.H.

    1977-01-01

    Some comparisons of transient fluid-structure tests with results from the ICECO containment code have already been published. The test data used before referred to simplifield structural models dealing with the safety aspect of the SNR-300 fast breeder. Thus, the comparison of these more complex models with the ICECO results is the subject of this paper. The experimental data used herein pertain to test models composed of a spherical source within a cylindrical pool water. The rigid ends of the container are connected by holddown bolts and cylindrical boundaries are made of one or two concentric deformable shells. The space above the surface of the water is occupied by air which may vent during the course of the experiment. The dimensions of the model, the transient stress-strain data of the shells, which had been derived in separate experiments, and the pressure-volume relationship are known. Although the boundary conditions of the vessel are also known, they could not be simulated by the analytical models. Initially the outside shell is prestressed by the holddown bolts through the two rigid lids with rubber seals set between the ends of the shell and the lids. This rather complex boundary condition was anlytically simulated in the following way: at the bottom the shells were assumed free to deform radially, but were fixed axially; at the top the shell was assumed to have no constraints at all. The analytical results show that the cylindrical vessel closest to the source begins to deform first, followed by the bottom portion of the vessel. The top vessel portion deforms only later when the fluid surface reaches the top cover and the developed fluid pressure imparts radial vessel deformation

  16. Pool-type reactor

    International Nuclear Information System (INIS)

    Hopkins, S.R.

    1977-01-01

    This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

  17. Recent development of a CEC'S elasto-plastic-creep cyclic benchmark programme relevant to LMFBR structural integrity

    International Nuclear Information System (INIS)

    Corsi, F.; Terzaghi, A.

    1984-01-01

    It's presented the programme of elasto-plastic benchmark calculations relevant to LMFBr, which started in 1977 with the support and coordination of the Commission of the European Communities (CEC) and the participation of nuclear engineering and manufacturing companies as well as nuclear research centers of France, Germany, Italy and the United Kingdom. (E.G.) [pt

  18. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    Magnaud, J.P.; Rouzaud, P.

    1985-09-01

    TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

  19. Effects of entrained gas on the acoustic detection of sodium boiling in a simulated LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Leavell, W.H.; Sides, W.H.

    1975-01-01

    The relationship between acoustic intensity of nucleate boiling and void fraction was studied in a simulated LMFBR fuel bundle. Results indicate that as the void fraction increases the detected intensity of nucleate boiling decreased until it was indistinguishable from background noise. (JWR)

  20. Model-based temperature noise monitoring methods for LMFBR core anomaly detection

    International Nuclear Information System (INIS)

    Tamaoki, Tetsuo; Sonoda, Yukio; Sato, Masuo; Takahashi, Ryoichi.

    1994-01-01

    Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an autoregressive model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the autoregressive model has a higher precision than the physical model in background noise prediction. Based on these results, an 'autoregressive model modification method' is proposed, in which a temporary autoregressive model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated autoregressive model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio. (author)

  1. Bulk coolant cavitation in LMFBR containment loading following a whole-core explosion

    International Nuclear Information System (INIS)

    Jones, A.V.

    1977-01-01

    An LMFBR core undergoing an explosion transmits energy to the containment in a series of pressure waves and the containment loading is determined by their cumulative effect. These pressure waves are modified by their interaction with the coolant through which they propagate. It is necessary to model both the induction of bulk cavitation by tension waves and the interaction of pressure waves with cavitated liquid in realistic containment loading calculations. This paper sets out the progress which has been achieved in such modelling and first indications for the effect of bulk coolant cavitation in LMFBR containment loading. Conclusions may be briefly summarised: 1) Bulk cavitation must be included in realistic containment loading calculations. 2) Phenomenological models of cavitated liquid without memory are inappropriate. The best approach is to model bubble dynamics directly, including at least momentum conservation and surface tension. 3) The containment loading resulting from a given explosion is sensitive to the state of preparation of the coolant. The number density of nucleation sites should therfore accompany the results of model tests. (Auth.)

  2. A study on reactor core failure thresholds to safety operation of LMFBR

    International Nuclear Information System (INIS)

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-01-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  3. International Atomic Energy Agency specialist meeting on advances in structural analysis for LMFBR applications. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, M A; Roche, R L [eds.

    1983-05-01

    After the first session on review of national positions in the subject field, the meeting was divided into five technical sections as follows: General methods of Structural Analysis for Elevated Temperatures; Inelastic Analysis Methods for Elevated Temperature; Effects of Cyclic loading; Design Codes and Criteria; Instability and Buckling - Piping Analysis in the Creep Range. The conclusions of the Meeting were summarised as follows. In view of the complexity of material behaviour and the variability of properties from cast to cast, continuing work is needed to develop simple constitutive relations which ensure an acceptable level of conservatism for design evaluations. It is recognized that simplified design methods require further development for the assessment of ratchetting and shakedown of high temperature structures. More development work is required in the areas of buckling elastic follow up weld factors and these developments should take account of the imperfections inherent in welded fabrications. There is a need for realistic tests on welded structural features to validate design methods. It is proposed that this subject would be the topic of a future specialists meeting. In several countries, organisations are now preparing Guides and Codes concerning Structural Assessment for LMFBR components. It seems that some of these Codes could be drafted within a few years. In order to make a more realistic assessment of LMFBR structures, defect assessment in elevated temperature range must be considered.

  4. International Atomic Energy Agency specialist meeting on advances in structural analysis for LMFBR applications. Summary report

    International Nuclear Information System (INIS)

    Perez, M.A.; Roche, R.L.

    1983-05-01

    After the first session on review of national positions in the subject field, the meeting was divided into five technical sections as follows: General methods of Structural Analysis for Elevated Temperatures; Inelastic Analysis Methods for Elevated Temperature; Effects of Cyclic loading; Design Codes and Criteria; Instability and Buckling - Piping Analysis in the Creep Range. The conclusions of the Meeting were summarised as follows. In view of the complexity of material behaviour and the variability of properties from cast to cast, continuing work is needed to develop simple constitutive relations which ensure an acceptable level of conservatism for design evaluations. It is recognized that simplified design methods require further development for the assessment of ratchetting and shakedown of high temperature structures. More development work is required in the areas of buckling elastic follow up weld factors and these developments should take account of the imperfections inherent in welded fabrications. There is a need for realistic tests on welded structural features to validate design methods. It is proposed that this subject would be the topic of a future specialists meeting. In several countries, organisations are now preparing Guides and Codes concerning Structural Assessment for LMFBR components. It seems that some of these Codes could be drafted within a few years. In order to make a more realistic assessment of LMFBR structures, defect assessment in elevated temperature range must be considered

  5. Power DRAC for rapid LMFBR deployment and consequent CO2 mitigation

    International Nuclear Information System (INIS)

    Schenewerk, W.E.

    2006-01-01

    A metallic-sodium LMFBR (Liquid Metal Fast Breeder Reactor) can control fuel temperature after a full power SCRAM using natural convection. A 3 percent nominal DRAC (Direct Reactor Auxiliary Cooling) does this without moving parts. DRAC is promoted from tertiary to primary decay heat removal, resulting in what is referred to as a Power DRAC. Power DRAC operates continuously before and after SCRAM, rejecting 3 per cent pile power. Power DRAC operability is validated by having it reject 75 MWt from a 2500 MWt pile at all times. IHX (Intermediate Heat Exchanger) is not required to be operable for primary, secondary, or tertiary core over temperature protection. Original DRAC concept (venturi DRAC) was a 1 per cent nominal tertiary decay heat removal system. Tertiary DRAC patent has expired. Power DRAC rejects 75 MWt through its own secondary sodium heat transfer loop to power a 25 MWe air Brayton cycle. Power DRAC eliminates requiring steam plant operability for decay heat removal. Intermediate sodium heat transfer system and steam plant can be optimized for maximum thermal efficiency. 2.5 GWt pile makes 1.0 GWe net power. Power DRAC maintains pile inlet and outlet temperatures while going from power to post-SCRAM conditions. Steam pressure is maintained post-SCRAM to mitigate SCRAM thermal transient. Not requiring steam plant operability for decay heat removal eases licensing and allows early LMFBR deployment. Each GWe atomic power delays Co2 doubling one week. (author)

  6. Main aspects of the design of a support structure of a LMFBR with particular reference to the explosive accident consequences

    International Nuclear Information System (INIS)

    Giuliano, V.; Lazzeri, L.

    1977-01-01

    The aim of this paper is a review of the main aspects of the design of a support structure of a LMFBR tank, with particular reference to the analysis of the non-linear dynamic behavior of the structure in the plastic range under the effect of an explosive accident within the tank. The structure is composed by a L-shaped flange, which supports the tank, connected by means of nine square beams to a rigid box-type ring, fixed to the concrete. The plug of the tank is connected to the L-shaped flange by means of a group of SS bars. The non-linear dynamic analysis of the explosive accident has been carried out on a lumped mass model, with elastic-plastic elements which simulate main components of the support structure and tank. The impulsive load connected to the explosive accident has been modelled (on the basis of extensive comparative studies carried out) as two triangular pressure impulses has been the object of a parametric evaluation. The dynamic transient on the support structure during and after the explosive accident for each couple of pressure impulses has been analyzed by means of modified version of the NON SAP code running on a CDC 7600 computer. A large amount of results, which describe displacements, velocities and accelerations of the plug, of the tank, and of the support structure, together with the forces and stresses acting on the main structural components are presented and discussed, with particular reference to the influence of the various parameters involved in the analysis

  7. Steam blowdown experiments with the condensation pool test rig

    International Nuclear Information System (INIS)

    Purhonen, H.; Puustinen, M.; Laine, J.; Raesaenen, A.; Kyrki-Rajamaeki, R.; Vihavainen, J.

    2005-01-01

    During a possible loss-of-coolant accident (Local) a large amount of non-condensable (nitrogen) and condensable (steam) gas is blown from the upper drywell of the containment to the condensation pool through the blowdown pipes at the boiling water reactors (BWRs). The wet well pool serves as the major heat sink for condensation of steam. The blowdown causes both dynamic and structural loads to the condensation pool. There might also be a risk that the gas discharging to the pool could push its way to the emergency core cooling systems (ECCS) and undermine their performance. (author)

  8. Study group meeting on steam generators for LMFBR's. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1975-07-01

    The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks.

  9. Study group meeting on steam generators for LMFBR's. Summary report

    International Nuclear Information System (INIS)

    1975-07-01

    The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks

  10. Ultrasonic scanner for stainless steel weld inspections. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kupperman, D. S.; Reimann, K. J.

    1978-09-01

    The large grain size and anisotropic nature of stainless steel weld metal make conventional ultrasonic testing very difficult. A technique is evaluated for minimizing the coherent ultrasonic noise in stainless steel weld metal. The method involves digitizing conventional ''A-scan'' traces and averaging them with a minicomputer. Results are presented for an ultrasonic scanner which interrogates a small volume of the weld metal while averaging the coherent ultrasonic noise.

  11. Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

    International Nuclear Information System (INIS)

    Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

    1976-01-01

    Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

  12. Air-cleaning devices for vented filtered LMFBR containment

    International Nuclear Information System (INIS)

    Muhlestein, L.D.; Hilliard, R.K.

    1982-07-01

    An effort lasting several years is summarized which evaluated, developed and tested air cleaning devices for potential use in breeder reactor containment venting applications. State-of-technology evaluations were completed for both a hypothetical head release accident and a primary vessel melt-through accident. Commercially available systems or components were tested which included HEPA filters, sand and gravel beds, and aqueous scrubbers. Large-scale demonstration tests were completed and results are presented for two- and three-stage conventional aqueous scrubber systems; and for a newly developed passive, submerged gravel scrubber

  13. Corium quench in deep pool mixing experiments

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.; Gregorash, D.; Aeschlimann, R.; Sienicki, J.J.

    1985-01-01

    The results of two recent corium-water thermal interaction (CWTI) tests are described in which a stream of molten corium was poured into a deep pool of water in order to determine the mixing behavior, the corium-to-water heat transfer rates, and the characteristic sizes of the quenched debris. The corium composition was 60% UO 2 , 16% ZrO 2 , and 24% stainless steel by weight; its initial temperature was 3080 K, approx.160 K above the oxide phase liquidus temperature. The corium pour stream was a single-phase 2.2 cm dia liquid column which entered the water pool in film boiling at approx.4 m/s. The water subcooling was 6 and 75C in the two tests. Test results showed that with low subcooling, rapid steam generation caused the pool to boil up into a high void fraction regime. In contrast, with large subcooling no net steam generation occurred, and the pool remained relatively quiescent. Breakup of the jet appeared to occur by surface stripping. In neither test was the breakup complete during transit through the 32 cm deep water pool, and molten corium channeled to the base where it formed a melt layer. The characteristic heat transfer rates measured 3.5 MJ/s and 2.7 MJ/s during the fall stage for small and large subcooling, respectively; during the initial stage of bed quench, the surface heat fluxes measured 2.4 MW/m 2 and 3.7 MW/m 2 , respectively. A small mass of particles was formed in each test, measuring typically 0.1 to 1 mm and 1 to 5 mm dia for the large and small subcooling conditions, respectively. 9 refs., 13 figs., 1 tab

  14. Mathematical modellings and computational methods for structural analysis of LMFBR's

    International Nuclear Information System (INIS)

    Liu, W.K.; Lam, D.

    1983-01-01

    In this paper, two aspects of nuclear reactor problems are discussed, modelling techniques and computational methods for large scale linear and nonlinear analyses of LMFBRs. For nonlinear fluid-structure interaction problem with large deformation, arbitrary Lagrangian-Eulerian description is applicable. For certain linear fluid-structure interaction problem, the structural response spectrum can be found via 'added mass' approach. In a sense, the fluid inertia is accounted by a mass matrix added to the structural mass. The fluid/structural modes of certain fluid-structure problem can be uncoupled to get the reduced added mass. The advantage of this approach is that it can account for the many repeated structures of nuclear reactor. In regard to nonlinear dynamic problem, the coupled nonlinear fluid-structure equations usually have to be solved by direct time integration. The computation can be very expensive and time consuming for nonlinear problems. Thus, it is desirable to optimize the accuracy and computation effort by using implicit-explicit mixed time integration method. (orig.)

  15. Microcomputer-controlled ultrasonic data acquisition system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, W.A. Jr.

    1978-11-01

    The large volume of ultrasonic data generated by computer-aided test procedures has necessitated the development of a mobile, high-speed data acquisition and storage system. This approach offers the decided advantage of on-site data collection and remote data processing. It also utilizes standard, commercially available ultrasonic instrumentation. This system is controlled by an Intel 8080A microprocessor. The MCS80-SDK microcomputer board was chosen, and magnetic tape is used as the storage medium. A detailed description is provided of both the hardware and software developed to interface the magnetic tape storage subsystem to Biomation 8100 and Biomation 805 waveform recorders. A boxcar integrator acquisition system is also described for use when signal averaging becomes necessary. Both assembly language and machine language listings are provided for the software.

  16. Intermediate-Size Inducer Pump design report. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, T.J.

    1979-06-15

    This report summarizes the mechanical, structural, and hydrodynamic design of the Intermediate-Size Inducer Pump (ISIP). The design was performed under Atomics International's DOE Base Technology Program by the Atomics International and Rocketdyne Divisions of Rockwell International. The pump was designed to utilize the FFTF prototype pump frame as a test vehicle to test the inducer, impeller, and diffuser plus necessary adapter hardware under simulated Large Scale Liquid Metal Fast Breeder Reactor service conditions. The report describes the design requirements including the purpose and objectives, and discusses those design efforts and considerations made to meet the requirements. Included in the report are appendices showing calculative methods and results. Also included are overall assembly and layout drawings plus some details used as illustrations for discussion of the design results and the results of water tests performed on a model of the inducer.

  17. Rotary plug device for use in LMFBR type reactors

    International Nuclear Information System (INIS)

    Azuma, Kazuhiko; Imayoshi, Sho.

    1988-01-01

    Purpose: To prevent adhesion of sodium in the rotational gap of a rotational plug. Constitution: One of the walls of a cylindrical gap formed between the outer circumference of a small rotary plug and a large rotary plug that constitute a double rotary plug is cooled to lower than the sodium coagulation temperature, while a stater of a linear motor in a cylindrical shape and wound with linear coils around the iron core is attached to the inside of the other of the walls. Then, one of the walls of the gap to which sodium adheres is cooled to less than sodium coagulation temperature, so that sodium is or tends to be deposited to the wall. Then, eddy currents are resulted to sodium by the current supplied to the stater of the linear motor attached to the other of the walls, to produce thrusting force. Sodium on the wall surface is scraped off by this. (Yoshihara, H.)

  18. Test system to simulate transient overpower LMFBR cladding failure

    International Nuclear Information System (INIS)

    Barrus, H.G.; Feigenbutz, L.V.

    1981-01-01

    One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

  19. Studies of spatial decoupling in heterogeneous LMFBR critical assemblies

    International Nuclear Information System (INIS)

    Brumbach, S.B.; Goin, R.W.; Carpenter, S.G.

    1984-01-01

    Recent measurements at the Zero Power Plutonium Reactor have studied the spatial decoupling in large, heterogeneous assemblies. These assemblies exhibited a significantly greater degree of decoupling than previous homogeneous assemblies of similar size. The flux distributions in these heterogeneous assemblies were very sensitive reactivity perturbations, and perturbed flux distributions were achieved relatively slowly. Decoupling was investigated using rod-drop, boron-oscillator and noise-coherence techniques which emphasized different times following the perturbations. Reactivity changes could be measured by analyzing the power history from a single detector using inverse kinetics methods with the assumption of an instantaneous efficiency change for the detector. For assemblies more decoupled than ZPPR-13, the instantaneous efficiency change assumption begins to be invalid

  20. Livestock Grazing as a Driver of Vernal Pool Ecohydrology

    Science.gov (United States)

    Michaels, J.; McCarten, N. F.

    2017-12-01

    Vernal pools are seasonal wetlands that host rare plant communities of high conservation priority. Plant community composition is largely driven by pool hydroperiod. A previous study found that vernal pools grazed by livestock had longer hydroperiods compared with pools excluded from grazing for 10 years, and suggests that livestock grazing can be used to protect plant diversity. It is important to assess whether observed differences are due to the grazing or due to water balance variables including upland discharge into or out of the pools since no a priori measurements were made of the hydrology prior to grazing. To address this question, in 2016 we compared 15 pools that have been grazed continuously and 15 pools that have been fenced off for over 40 years at a site in Sacramento County. We paired pools based on abiotic characteristics (size, shape, slope, soil type) to minimize natural variation. We sampled vegetation and water depth using Solinst level loggers. We found that plant diversity and average hydroperiod was significantly higher in the grazed pools. We are currently measuring groundwater connectivity and upland inputs in order to compare the relative strength of livestock grazing as a driver of hydroperiod to these other drivers.

  1. On the computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.-W.; Fistedis, S.H.

    1977-01-01

    A two-dimensional coupled hydrodynamic-structural response analysis of piping systems is described. Implicit Continuous-Fluid Eulerian (ICE) technique is utilized in the hydrodynamics while a finite-element technique is used in the structural analysis. Different piping components such as elbows, valves, reducers, expansions, heat exchangers, and tees are modelled and coupled with the straight pipe model. An axisymmetric general component model that can be used in modelling valves, reducers, expansions, and heat exchangers is described. At the inlet and outlet region of such component the cross-sectional area may change suddently or gradually, or many not change at all. Among the options available in this model are deformable exterior walls, interior rigid wall simulation, and tube bundle effect. Exterior walls of pipes and components are treated as thin axisymmetric shell. A convected coordinate explicit finite-element scheme for large displacement small strain, elastic-plastic material behavior in which membrane and bending strengths are accounted for is employed. The strains are linearly related to the displacement of the element relative to its convective coordinates, and similarly, the nodal forces are linearly related to the elements stresses. The coupling of the hydrodynamics and structural problems is done in such a way that the hydrodynamics supplies the structure with a pressure loading and the structure supplying the hydrodynamics with a moving boundary condition. Because of the difficulties of handling interior walls that may occupy partial zones, the walls are assumed rigid and limited in their orientation to be parallel to the radial or axial directions, their position to zone boundaries, and their thickness to zero

  2. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  3. Analysis of pressure wave transients and seismic response in LMFBR piping systems using the SHAPS code

    International Nuclear Information System (INIS)

    Zeuch, W.R.; Wang, C.Y.

    1985-01-01

    This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs

  4. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  5. Sodium-NaK engineering handbook. Volume III. Sodium systems, safety, handling, and instrumentation. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Foust, O J [ed.

    1978-01-01

    The handbook is intended for use by present and future designers in the Liquid Metals Fast Breeder Reactor (LMFBR) Program and by the engineering and scientific community performing other type investigation and exprimentation requiring high-temperature sodium and NaK technology. The arrangement of subject matter progresses from a technological discussion of sodium and sodium--potassium alloy (NaK) to discussions of varius categories and uses of hardware in sodium and NaK systems. Emphasis is placed on sodium and NaK as heat-transport media. Sufficient detail is included for basic understanding of sodium and NaK technology and of technical aspects of sodium and NaK components and instrument systems. Information presented is considered adequate for use in feasibility studies and conceptual design, sizing components and systems, developing preliminary component and system descriptions, identifying technological limitations and problem areas, and defining basic constraints and parameters.

  6. Key technological issues in LMFBR high-temperature structural design - the US perspective

    International Nuclear Information System (INIS)

    Corum, J.M.

    1984-01-01

    The purpose of this paper is: (1) to review the key technological issues in LMFBR high-temperature structural design, particularly as they relate to cost reduction; and (2) to provide an overview of activities sponsored by the US Department of Energy to resolve the issues and to establish stable, standardized, and defensible structural design methods and criteria. Specific areas of discussion include: weldments, structural validation tests, simplified design analysis procedures, design procedures for piping, validation of the methodology for notch-like geometries, improved life assessment procedures, thermal striping, extension of the methodology to new materials, and ASME high-temperature Code reform needs. The perceived problems and needs in each area are discussed, and the current status of related US activities is given

  7. Status of the LMFBR thermo- and fluid-dynamic activities at KFK

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hofmann, F.; Rehme, K.

    1979-01-01

    The aim of the thermo- and fluiddynamic analysis is to determine the spatial velocity and temperature distributions in LMFBR-core elements with high accuracy. Knowledge of these data is a necessary prerequisite for determining the mechanical behavior of fuel rods and of structural material. Three cases are distinguished: Nominal geometry and steady state conditions; non-nominal geometry and quasi-steady state conditions; nominal geometry and non-steady state conditions. The present situation for the design calculations of fuel elements is based mainly on undisturbed normal operation. Most of the thermo- and fluiddynamic activities performed under the Fast Breeder Programme at KFK are related to this case. The present status of theoretical and experimental research work briefly presented in this paper, can be subdivided into the following main topics: 1. Physical and mathematical modelling of single phase rod bundle thermo- and fluiddynamics, 2. Experimental investigations on heat transfer and fluid flow in rod bundles

  8. Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

    1979-01-01

    This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

  9. Active acoustic leak detection for LMFBR steam generator. Sound attenuation due to bubbles

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Sakuma, Toshio

    1995-01-01

    In the steam generators (SG) of LMFBR, it is necessary to detect the leakage of water from tubes of heat exchangers as soon as it occurs. The active acoustic detection method has drawn general interest owing to its short response time and reduction of the influence of background noise. In this paper, the application of the active acoustic detection method for SG is proposed, and sound attenuation by bubbles is investigated experimentally. Furthermore, using the SG sector model, sound field characteristics and sound attenuation characteristics due to injection of bubbles are studied. It is clarified that the sound attenuation depends upon bubble size as well as void fraction, that the distance attenuation of sound in the SG model containing heat transfer tubes is 6dB for each two-fold increase of distance, and that emitted sound attenuates immediately upon injection of bubbles. (author)

  10. Collection and evaluation of salt mixing data with the real time data acquisition system. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Glazer, S.; Chiu, C.; Todreas, N.E.

    1977-09-01

    A minicomputer based real time data acquisition system was designed and built to facilitate data collection during salt mixing tests in mock ups of LMFBR rod bundles. The system represents an expansion of data collection capabilities over previous equipment. It performs steady state and transient monitoring and recording of up to 512 individual electrical resistance probes. Extensive real time software was written to govern all phases of the data collection procedure, including probe definition, probe calibration, salt mixing test data acquisition and storage, and data editing. Offline software was also written to permit data examination and reduction to dimensionless salt concentration maps. Finally, the computer program SUPERENERGY was modified to permit rapid extraction of parameters from dimensionless salt concentration maps. The document describes the computer system, and includes circuit diagrams of all custom built components. It also includes descriptions and listings of all software written, as well as extensive user instructions.

  11. Material properties requirements for LMFBR structural design: General considerations and data needs

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C E [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Purdy, C M [U.S. Energy Research and Development Administration (United States)

    1977-07-01

    A statement is given of material properties information needed in connection with the structural design technology for liquid-metal fast breeder reactor (LMFBR) primary circuit components. Implementation of current analysis methods and criteria is considered with an emphasis on data and data correlations for performing elastic-plastic and creep analyses, for establishing allowable stress limits, and for computing creep-fatigue damage. Further development of the technology is discussed in relation to properties information. Emphasis is placed on improved constitutive equations for representing inelastic material behavior, on procedures for treating time-dependent fatigue, and on criteria for creep rupture. The properties are generally discussed without regard to specific alloys, since most categories of information are needed for each major structural material. Some sample experimental results are given for type 304 stainless steel and 2 1/4 Cr-1 Mo steel. (author)

  12. Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

    International Nuclear Information System (INIS)

    Wong, C.N.C.; Todreas, N.E.

    1982-07-01

    Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

  13. Finite element analysis of irradiation-induced dilation of the fuel subassembly duct in LMFBR

    International Nuclear Information System (INIS)

    Gao Fuhai; Fu Hao; Li Nan; Yang Kongli; Wang Mingzhen

    2013-01-01

    Background: The calculation of irradiation-induced dilation of the fuel subassembly duct in LMFBR is important for fast reactor core design.. Purpose: To investigate how to calculate the dilation by using finite element method (FEM). Methods: First, irradiation-induced creep and swelling material models are introduced. Then, a theoretical solution based on a simplified bending plate model is briefly given. Finally, a stress update scheme for the adopted material models is presented and furthermore embedded into ABAQUS user interface UMAT to conduct finite element analysis. Both solutions are compared and discussed. Results: FEM successfully predicts the duct dilation and its solution agrees well with theoretical one in small deformation. Conclusions: The proposed stress update scheme is effective, The accuracy of the theory solution declines when dilation becomes larger. The maximum stress occurs at the duct corner point, and the location has stress relaxation effect. (authors)

  14. Safety evaluation for the LMFBR plant using probabilistic risk assessment techniques

    International Nuclear Information System (INIS)

    Kani, Y.; Aizawa, K.

    1987-01-01

    This paper presents an application of probabilistic risk assessment techniques to a typical loop-type liquid metal fast breeder reactor (LMFBR) plant in the detailed design stage. A comprehensive systems analysis has been performed to identify event sequences leading to core damage and provide insights into the importance of accident contributors. While traditional event tree/fault tree modeling was used for the analysis, this study involved a thorough investigation of initiating events and of support system faults. The qualification of accident sequences has been conducted by combining the fault trees based on the event trees and obtaining sequence cut sets with the use of the SETS code. This study also attempted to quantify the potential for operator recovery actions in the course of each accident sequence. (author)

  15. Single-phase pump model for analysis of LMFBR heat transport systems

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.

    1978-05-01

    A single-phase pump model for transient and steady-state analysis of LMFBR heat transport systems is presented. Fundamental equations of the model are angular momentum balance to determine transient impeller speed and mass balance (including thermal expansion effects) to determine the level of sodium in the pump tank. Pump characteristics are modeled by homologous head and torque relations. All regions of pump operation are represented with reverse rotation allowed. The model also includes option for enthalpy rise calculations and pony motor operation. During steady state, the pump operating speed is determined by matching required head with total load in the circuit. Calculated transient results are presented for pump coastdown and double-ended pipe break accidents. The report examines the influence of frictional torque and specific speed on predicted response for the pump coastdown to natural circulation transient. The results for a double-ended pipe break accident indicate the necessity of including all regions of operation for pump characteristics

  16. Material properties requirements for LMFBR structural design: general considerations and data needs

    International Nuclear Information System (INIS)

    Pugh, C.E.; Purdy, C.M.

    1977-01-01

    A statement is given of material properties information needed in connection with the structural design technology for liquid-metal fast breeder reactor (LMFBR) primary circuit components. Implementation of current analysis methods and criteria is considered with an emphasis on data and data correlations for performing elastic-plastic and creep analyses, for establishing allowable stress limits, and for computing creep-fatigue damage. Further development of the technology is discussed in relation to properties information. Emphasis is placed on improved constitutive equations for representing inelastic material behavior, on procedures for treating time-dependent fatigue, and on criteria for creep rupture. The properties are generally discussed without regard to specific alloys, since most categories of information are needed for each major structural material. Some sample experimental results are given for type 304 stainless steel and 2 1 / 4 Cr-1 Mo steel

  17. TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

    International Nuclear Information System (INIS)

    Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

    1981-01-01

    The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

  18. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  19. Laminar/transition sweeping flow-mixing model for wire-wrapped LMFBR assemblies

    International Nuclear Information System (INIS)

    Burns, K.F.; Rohsenow, W.M.; Todreas, N.E.

    1980-07-01

    Recent interest in analyzing the thermal hydraulic characteristics of LMFBR assemblies operating in the mixed convection regime motivates the extension of the aforementioned turbulent sweeping flow model to low Reynolds number flows. The accuracy to which knowledge of the mixing parameters is required has not been well determined, due to the increased influence of conduction and buoyancy effects with respect to energy transport at low Reynolds numbers. This study represents a best estimate attempt to correlate the existing low Reynolds number sweeping flow data. The laminar/transition model which is presented is expected to be useful in anayzing mixed convection conditions. However, the justification for making additional improvemements is contingent upon two factors. First, the ability of the proposed laminar/transition model to predict additional low Reynolds number sweeping flow data for other geometries needs to be investigated. Secondly, the sensitivity of temperature predictions to uncertainties in the values of the sweeping flow parameters should be quantified

  20. Analytical model for transient fluid mixing in upper outlet plenum of an LMFBR

    International Nuclear Information System (INIS)

    Yang, J.W.; Agrawal, A.K.

    1976-01-01

    A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the outlet plenum of an LMFBR. The maximum penetration of core flow is used as the criterion for dividing the sodium region into two mixing zones. The model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of by-pass flow into the plenum. The results of numerical calculations indicate that effects of flow stratification, chimney height, metal heat capacity and by-pass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas do not play any significant role on sodium temperature

  1. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    International Nuclear Information System (INIS)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified

  2. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  3. Materials properties utilization in a cumulative mechanical damage function for LMFBR fuel pin failure analysis

    International Nuclear Information System (INIS)

    Jacobs, D.C.

    1977-01-01

    An overview is presented of one of the fuel-pin analysis techniques used in the CRBRP program, the cumulative mechanical damage function. This technique, as applied to LMFBR's, was developed along with the majority of models used to describe the mechanical properties and environmental behavior of the cladding (i.e., 20 percent cold-worked, 316 stainless steel). As it relates to fuel-pin analyses the Cumulative Mechanical Damage Function (CDF) continually monitors cladding integrity through steady state and transient operation; it is a time dependent function of temperature and stress which reflects the effects of both the prior mechanical history and the variations in mechanical properties caused by exposure to the reactor environment

  4. Structural consideration for hot and cold pipe clamps in LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.; Huang, S.N.; Kappauf, H.; Wagner, S.E.; Wirtz, K.H.

    1983-01-01

    A series of analytical studies are described which evaluate stress levels induced in a 600 mm high temperature, thin-wall sodium pipeline by two alternate clamp designs. The first design consists of a band mounted directly on the pipe and is called the hot clamp. The second design consists of a band mounted using insulation standoffs and is called the cold clamp. Pipe stress levels induced by transient thermal dead weight and seismic loads are discussed. Pipe stress levels and system dynamic spring rates are presented. Procedures utilized to combine clamp induced pipe stress with other short and long term pipe system stresses are detailed. Recommendations for practical application in LMFBR pipe systems are made

  5. Structural considerations for hot and cold pipe clamps in LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.; Huang, S.N.; Wagner, S.E.; Kappauf, H.; Wirtz, K.H.

    1983-01-01

    A series of analytical studies are described which evaluate stress levels induced in a 600 mm high temperature, thin-wall sodium pipeline by two alternate clamp designs. The first design consists of a band mounted directly on the pipe and is called the hot clamp. The second design consists of a band mounted using insulation standoffs and is called the cold clamp. Pipe stress levels induced by transient thermal dead weight and seismic loads are discussed. Pipe stress levels and system dynamic spring rates are presented. Procedures utilized to combine clamp induced pipe stress with other short and long term pipe system stresses are detailed. Recommendations for practical application in LMFBR pipe systems are made

  6. Research report on design allowable values of structural materials for LMFBR

    International Nuclear Information System (INIS)

    1978-11-01

    The present report is composed of following two main parts. i) review and re-evaluation on test results by FCI Sub-committee studies, performed from 1973 to 1976, ii) review on procedures for determining design allowable values of structural materials for LMFBR components. Re-evaluation works have been made on monotonic tensile properties at elevated temperatures, creep and creep rupture properties, creep-fatigue properties (strain rate and tensile strain hold time effects on strain fatigue properties at elevated temperatures) of Types 316 and 304 stainless steel and 2 1/4Cr-1Mo steel (base and weld metals) produced in Japan. In the first half of the present report, creep-fatigue test results obtained by FCI Sub-committee studies are subjected to re-evaluation by the present P-FCI Sub-committee. Reviews have been made on testing methods on FCI's-creep-fatigue experiments with other test data of the test materials; high temperature monotonic tensile data, creep and creep rupture data, and origin of the test materials. The data of FCI studies are compared with other reference data obtained by several Japanese laboratories. In the latter half of the present report, procedures including ASME's are reviewed for setting design allowable values for LMFBR components on the basis of high temperature strength properties obtained with materials produced in Japan. A creep rupture data of Japanese steels are issued and examined to make proposal for a design allowable stress of S sub(t) through parameter survey. (author)

  7. Structural dynamics in LMFBR containment analysis. A brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.

    1977-01-01

    This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers to up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. First, the commonly used finite-difference expressions in the Lagrangian codes will be compared. Sample calculations will be used as a basis for discussing and comparing the accuracy of the various finite-difference representations. The distortion of the meshes will also be compared; the techniques used for eliminating the numerical instabilities will be discussed and compared using examples. Next, the numerical methods used in the Eulerian formulation will be compared, first among themselves and then with the Lagrangian formulations. Special emphasis is placed on the effect of mass diffusion of the Eulerian calculation on the propagation of discontinuities. Implicit and explicit numerical integrations will be discussed and results obtained from these two techniques will be compared. Then, the finite-element methods are compared with the finite-difference methods. The advantages and disadvantages of the two methods will be discussed in detail, together with the versatility and ease of application of the method to containment analysis having complex geometries. It will also be shown that the finite-element equations for a constant-pressure fluid element is identical to the finite-difference equations using contour integrations. Finally, conclusions based on this study will be given

  8. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  9. Design of hydrotherapy exercise pools.

    Science.gov (United States)

    Edlich, R F; Abidin, M R; Becker, D G; Pavlovich, L J; Dang, M T

    1988-01-01

    Several hydrotherapy pools have been designed specifically for a variety of aquatic exercise. Aqua-Ark positions the exerciser in the center of the pool for deep-water exercise. Aqua-Trex is a shallow underwater treadmill system for water walking or jogging. Swim-Ex generates an adjustable laminar flow that permits swimming without turning. Musculoskeletal conditioning can be accomplished in the above-ground Arjo shallow-water exercise pool. A hydrotherapy pool also can be custom designed for musculoskeletal conditioning in its shallow part and cardiovascular conditioning in a deeper portion of the pool. Regardless of the type of exercise, there is general agreement that the specific exercise conducted in water requires significantly more energy expenditure than when the same exercise is performed on land.

  10. Swimming pools as heat sinks for air conditioners: Model design and experimental validation for natural thermal behavior of the pool

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Jonathan; Harrington, Curtis; Modera, Mark [University of California Davis, Western Cooling Efficiency Center, 1450 Drew Avenue, Suite 100, Davis, CA 95618 (United States)

    2011-01-15

    Swimming pools as thermal sinks for air conditioners could save approximately 40% on peak cooling power and 30% of overall cooling energy, compared to standard residential air conditioning. Heat dissipation from pools in semi-arid climates with large diurnal temperature shifts is such that pool heating and space cooling may occur concurrently; in which case heat rejected from cooling equipment could directly displace pool heating energy, while also improving space cooling efficiency. The performance of such a system relies on the natural temperature regulation of swimming pools governed by evaporative and convective heat exchange with the air, radiative heat exchange with the sky, and conductive heat exchange with the ground. This paper describes and validates a model that uses meteorological data to accurately predict the hourly temperature of a swimming pool to within 1.1 C maximum error over the period of observation. A thorough review of literature guided our choice of the most appropriate set of equations to describe the natural mass and energy exchange between a swimming pool and the environment. Monitoring of a pool in Davis, CA, was used to confirm the resulting simulations. Comparison of predicted and observed pool temperature for all hours over a 56 day experimental period shows an R-squared relatedness of 0.967. (author)

  11. The radiological significance of transuranium radioisotopes released to the environment during operation of the LMFBR fuel cycle

    International Nuclear Information System (INIS)

    Barr, N.F.

    1976-01-01

    Estimates based on current knowledge and conservative assumptions indicate that release of transuranium elements from the Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycle are likely to proaduce population dose commitments small compared to those produced by naturally occurring alpha emitters and globally dispersed transuranium radioisotopes from tests of nuclear weapons in the atmosphere. Potential health consequences of these releases to current and future generations are estimated to be very small compared to risks associated with the production of energy by fossil fuels. The estimates are subject to a number of uncertainties imposed by lack of knowledge. Some of the uncertainties are not likely to be greatly reduced until LMFBR facilities are designed and operated. Others may be significantly reduced prior to facility design and operation. The paper discusses the sensitivity of the estimates to uncertainties and approches to reducing those uncertainties that strongly influence the estimates. (author)

  12. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    International Nuclear Information System (INIS)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist

  13. Detailed design consideration on wire-spaced LMFBR fuel subassemblies under the effects of uncertainties and non-nominal geometries

    International Nuclear Information System (INIS)

    Hishida, H.

    1979-01-01

    This paper explains some analytical methods for evaluating the effects of deviation in subchannel coolant flow rate from the nominal value due to fuel pin bundle deflection and manufacturing tolerances and of inter-sub-channel coolant mixing and local temperature rise due to a wire-spacer on the hot spot temperature. Numerical results are given in each chapter with respect to a prototype LMFBR core. (author)

  14. Trip report: United States LMFBR Steam Generator Team. IAEA symposium, Bensberg, Germany, October 14--17, 1974

    International Nuclear Information System (INIS)

    1974-01-01

    Information is presented concerning steam generator design characteristics for the AFR reactor, SNR reactor, PHENIX reactor, SUPER PHENIX reactor, MONJU reactor, and BN-350 reactor; steam generator development programs for West Germany, France, Japan, U. K., and the U. S. S. R.; and the fabrication and inspection of steam generator components. Steam generator performance and maintenance requirements for operating LMFBR reactors are reviewed. (U.S.)

  15. Tax credits and purchasing pools: will this marriage work?

    Science.gov (United States)

    Trude, S; Ginsburg, P B

    2001-04-01

    Bipartisan interest is growing in Congress for using federal tax credits to help low-income families buy health insurance. Regardless of the approach taken, tax credit policies must address risk selection issues to ensure coverage for the chronically ill. Proposals that link tax credits to purchasing pools would avoid risk selection by grouping risks similar to the way large employers do. Voluntary purchasing pools have had only limited success, however. This Issue Brief discusses linking tax credits to purchasing pools. It uses information from the Center for Studying Health System Change's (HSC) site visits to 12 communities as well as other research to assess the role of purchasing pools nationwide and the key issues and implications of linking tax credits and pools.

  16. poolHiTS: A Shifted Transversal Design based pooling strategy for high-throughput drug screening

    Directory of Open Access Journals (Sweden)

    Woolf Peter J

    2008-05-01

    Full Text Available Abstract Background A key goal of drug discovery is to increase the throughput of small molecule screens without sacrificing screening accuracy. High-throughput screening (HTS in drug discovery involves testing a large number of compounds in a biological assay to identify active compounds. Normally, molecules from a large compound library are tested individually to identify the activity of each molecule. Usually a small number of compounds are found to be active, however the presence of false positive and negative testing errors suggests that this one-drug one-assay screening strategy can be significantly improved. Pooling designs are testing schemes that test mixtures of compounds in each assay, thereby generating a screen of the whole compound library in fewer tests. By repeatedly testing compounds in different combinations, pooling designs also allow for error-correction. These pooled designs, for specific experiment parameters, can be simply and efficiently created using the Shifted Transversal Design (STD pooling algorithm. However, drug screening contains a number of key constraints that require specific modifications if this pooling approach is to be useful for practical screen designs. Results In this paper, we introduce a pooling strategy called poolHiTS (Pooled High-Throughput Screening which is based on the STD algorithm. In poolHiTS, we implement a limit on the number of compounds that can be mixed in a single assay. In addition, we show that the STD-based pooling strategy is limited in the error-correction that it can achieve. Due to the mixing constraint, we show that it is more efficient to split a large library into smaller blocks of compounds, which are then tested using an optimized strategy repeated for each block. We package the optimal block selection algorithm into poolHiTS. The MATLAB codes for the poolHiTS algorithm and the corresponding decoding strategy are also provided. Conclusion We have produced a practical version

  17. Macroinvertebrate community assembly in pools created during peatland restoration.

    Science.gov (United States)

    Brown, Lee E; Ramchunder, Sorain J; Beadle, Jeannie M; Holden, Joseph

    2016-11-01

    providing extensive new habitat that is largely equivalent to natural pools. More generally, we suggest that assembly theory could provide new benchmarks for planning and evaluating ecological restoration success. Copyright © 2016 The Authors. Published by Elsevier B.V. All rights reserved.

  18. Controls on the size and occurrence of pools in coarse-grained forest rivers

    Science.gov (United States)

    John M. Buffington; Thomas E. Lisle; Richard D. Woodsmith; Sue Hilton

    2002-01-01

    Controls on pool formation are examined in gravel- and cobble-bed rivers in forest mountain drainage basins of northern California, southern Oregon, and southeastern Alaska. We demonstrate that the majority of pools at our study sites are formed by flow obstructions and that pool geometry and frequency largely depend on obstruction characteristics (size, type, and...

  19. ENERGY STAR Certified Pool Pumps

    Data.gov (United States)

    U.S. Environmental Protection Agency — Certified models meet all ENERGY STAR requirements as listed in the Version 1.1 ENERGY STAR Program Requirements for Pool Pumps that are effective as of February 15,...

  20. Risk of HPV-16/18 Infections and Associated Cervical Abnormalities in Women Seropositive for Naturally Acquired Antibodies: Pooled Analysis Based on Control Arms of Two Large Clinical Trials.

    Science.gov (United States)

    Safaeian, Mahboobeh; Castellsagué, Xavier; Hildesheim, Allan; Wacholder, Sholom; Schiffman, Mark H; Bozonnat, Marie-Cécile; Baril, Laurence; Rosillon, Dominique

    2018-06-05

    Studies on the role of antibodies produced after infection with human papillomavirus 18 (HPV-18) and subsequent protection from HPV-18 infection have been conflicting, mainly due to inadequate sample size. We pooled data from the control arms of the Costa Rica Vaccine Trial and the PATRICIA trial. Using Poisson regression we compared the risk of newly detected 1-time HPV-18 infection, HPV-18 1-year persistent infection (12MPI), and HPV-18-associated atypical squamous cells of undetermined significance or greater (ASC-US+) lesions between HPV-18 seropositive and seronegative women. High HPV-18 antibodies at enrollment was associated with reduced subsequent HPV-18 detection (P trend = 0.001; relative rate [RR] = 0.69; 95% confidence interval [CI], 0.47-1.01 for the third quartile; RR = 0.63; 95% CI, 0.43-0.94 for the fourth quartile, compared to seronegative). The risk of 12MPI showed a decreasing trend with increasing antibodies (P trend = 0.06; RR = 0.72; 95% CI, 0.29-1.77; RR = 0.42; 95% CI, 0.13-1.32 for the third and fourth quartiles, respectively). Lastly, we observed a significant decreased risk of HPV-18 ASC-US+ with increasing antibody (P trend = 0.01; RR = 0.46; 95% CI, 0.21-0.97 for the fourth quartile). We also observed a significant decreased risk of HPV-16 infection, 12MPI, and ASC-US+ with increasing HPV-16 antibody level. High HPV-18 naturally acquired antibodies were associated with partial protection from future HPV-18 infections and associated lesions. NCT00128661 and NCT001226810.

  1. Grundfoss: Chlorination of Swimming Pools

    DEFF Research Database (Denmark)

    Hjorth, Poul G.; Hogan, John; Andreassen, Viggo

    1998-01-01

    Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools.......Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools....

  2. Sustainability of common pool resources

    OpenAIRE

    Timilsina, Raja Rajendra; Kotani, Koji; Kamijo, Yoshio

    2017-01-01

    Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as "capitalism," affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepales...

  3. Evaluation of Decontamination Factor of Aerosol in Pool Scrubber according to Bubble Shape and Size

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Hyun Joung; Ha, Kwang Soon; Jang, Dong Soon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The scrubbing pool could play an important role in the wet type FCVS because a large amount of aerosol is captured in the water pool. The pool scrubbing phenomena have been modelled and embedded in several computer codes, such as SPARC (Suppression Pool Aerosol Removal Code), BUSCA (BUbble Scrubbing Algorithm) and SUPRA (Suppression Pool Retention Analysis). These codes aim at simulating the pool scrubbing process and estimating the decontamination factors (DFs) of the radioactive aerosol and iodine gas in the water pool, which is defined as the ratio of initial mass of the specific radioactive material to final massy after passing through the water pool. The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the pool. The developed code has been verified using the experimental results and parametric studies the decontamination factor according to bubble shape and size. To evaluate the decontamination factor more accurate whole pool scrubber phenomena, the code was improved to consider the variety shape and size of bubbles. The decontamination factor were largely evaluated in ellipsoid bubble rather than in sphere bubble. The pool scrubbing models will be enhanced to apply more various model such as aerosol condensation of hygroscopic. And, it is need to experiment to measure to bubble shape and size distribution in pool to improve bubble model.

  4. A simple steel/water model for preliminary studies of acoustic vibration in LMFBR

    International Nuclear Information System (INIS)

    Bentley, P.G.; Firth, D.; Rowley, R.; Beesley, M.

    1977-01-01

    One source of vibration excitation in Liquid Metal Fast Breeder Reactors is the acoustic energy which is generated by the circulating pump and transmitted through the fluid to various structural components. Since most of the energy occurs at fairly low frequencies, that of low harmonies of blade passing frequency, only the very large components have resonant frequencies such that they are significantly excited. To gain some preliminary understanding of the extent and magnitude of vibration in fast reactors therefore, a simple model has been constructed in which only the major components are represented. The modelling theory is discussed and it is shown that adequate representation of the steel/sodium reactor materials can be obtained in the model based on the use of steel/water. The model represents a pool design with a primary tank of 3 1/4 metres diameter and typical components scaled in proportion; however, it does not necessarily relate to any specific reactor design. The pump acoustic source is represented by an underwater loudspeaker system and vibration amplitudes are scaled according to typical pressures generated by reactor circulators. Results from the model include calibration data for the acoustic source and measurements of acoustic pressure throughout the primary flow circuit and the inner and outer pools. Stresses are measured on structural components over a frequency range scaled from reactor frequencies and compensated for the characteristics of the acoustic source. Appreciable stresses are found on all the components in the primary circuit, not necessarily only those close to the simulated pump source. After scaling them to reactor size and allowing for the source calibration, it is found that stresses are unlikely to be sufficiently high to cause damage

  5. A swimming pool array for ultra high energy showers

    Science.gov (United States)

    Yodh, Gaurang B.; Shoup, Anthony; Barwick, Steve; Goodman, Jordan A.

    1992-11-01

    A very preliminary design concept for an array using water Cherenkov counters, built out of commercially available backyard swimming pools, to sample the electromagnetic and muonic components of ultra high energy showers at large lateral distances is presented. The expected performance of the pools is estimated using the observed lateral distributions by scintillator and water Cherenkov arrays at energies above 1019 eV and simulations.

  6. SNR-steam generator design with respect to large sodium water reactions

    International Nuclear Information System (INIS)

    Jong, J.J. de; Kellner, A.; Florie, C.J.L.

    1984-01-01

    This paper deals with the experiences gained during the licensing procedure for the steam generators for the SNR 300 LMFBR regarding large sodium-water reactions. A description is given of the different calculations executed to investigate the effects of large leaks on the 85 MW helical coiled and straight tube steam generators. The investigations on the helical coiled steam generators are divided in the formulations of fluid behaviour, dynamic force calculations, dynamic response calculation and finally stress analyses. Several results are shown. The investigations on the straight tube steam generators are performed using models describing fluid-structure interaction, coupled with stress analyses. Several results are presented. A description is given of the problems and necessary construction changes during the licensing process. Advises are given for future analyses and design concepts for second generation commercial size LMFBR steam generators with respect to large leaks; based on the experience, gained with SNR 300, and using some new calculations for SNR 2. (author)

  7. Controls on Filling and Evacuation of Sediment in Waterfall Plunge Pools

    Science.gov (United States)

    Scheingross, J. S.; Lamb, M. P.

    2014-12-01

    Many waterfalls are characterized by the presence of deep plunge pools that experience periods of sediment fill and evacuation. These cycles of sediment fill are a first order control on the relative magnitude of lateral versus vertical erosion at the base of waterfalls, as vertical incision requires cover-free plunge pools to expose the bedrock floor, while lateral erosion can occur when pools are partially filled and plunge-pool walls are exposed. Currently, there exists no mechanistic model describing sediment transport through waterfall plunge pools, limiting our ability to predict waterfall retreat. To address this knowledge gap, we performed detailed laboratory experiments measuring plunge-pool sediment transport capacity (Qsc_pool) under varying waterfall and plunge-pool geometries, flow hydraulics, and sediment size. Our experimental plunge-pool sediment transport capacity measurements match well with a mechanistic model we developed which combines existing waterfall jet theory with a modified Rouse profile to predict sediment transport capacity as a function of water discharge and suspended sediment concentration at the plunge-pool lip. Comparing the transport capacity of plunge pools to lower gradient portions of rivers (Qsc_river) shows that, for transport limited conditions, plunge pools fill with sediment under modest water discharges when Qsc_river > Qsc_pool, and empty to bedrock under high discharges when Qsc_pool > Qsc_river. These results are consistent with field observations of sand-filled plunge pools with downstream boulder rims, implying filling and excavation of plunge pools over single-storm timescales. Thus, partial filling of waterfall plunge pools may provide a mechanism to promote lateral undercutting and retreat of waterfalls in homogeneous rock in which plunge-pool vertical incision occurs during brief large floods that expose bedrock, whereas lateral erosion may prevail during smaller events.

  8. Gene pool conservation of teak in Myanmar

    International Nuclear Information System (INIS)

    Tin-Tun

    1995-01-01

    Myanmar with an area of 261, 228 Sq. miles is endowed with various types of forests which occupied nearly 50% of the country. Teak (Tectona grandis Linn. f.) is one of the most valuable timber species for its excellent wood quality and properties which are not observed with other timbers. Gene pool can be defined as a group of individual trees growing over a wide range of environmental conditions, and constituting different genetic complexes which can be transmitted to the offsprings. Topics such as: objectives of gene pool conservation, genetically improved seeds for large scale forest plantations, methodology of conservation, are discussed in the article. Myanmar teak dominates the world's teak market, and thus it is crucial to maintain the superiority in the conservation of gene complexes of teak. To some extent, the conservation of gene pools of teak and tree improvements are being undertaken by the Forest Research Institute of Myanmar. It is felt that the dissemination of the philosophy and concept of gene conservation to the personal involved in the forestry activities of the country are still inadequate

  9. Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

  10. A Pool of Distant Galaxies

    Science.gov (United States)

    2008-11-01

    Anyone who has wondered what it might be like to dive into a pool of millions of distant galaxies of different shapes and colours, will enjoy the latest image released by ESO. Obtained in part with the Very Large Telescope, the image is the deepest ground-based U-band image of the Universe ever obtained. It contains more than 27 million pixels and is the result of 55 hours of observations with the VIMOS instrument. A Sea of Galaxies ESO PR Photo 39/08 A Pool of Distant Galaxies This uniquely beautiful patchwork image, with its myriad of brightly coloured galaxies, shows the Chandra Deep Field South (CDF-S), arguably the most observed and best studied region in the entire sky. The CDF-S is one of the two regions selected as part of the Great Observatories Origins Deep Survey (GOODS), an effort of the worldwide astronomical community that unites the deepest observations from ground- and space-based facilities at all wavelengths from X-ray to radio. Its primary purpose is to provide astronomers with the most sensitive census of the distant Universe to assist in their study of the formation and evolution of galaxies. The new image released by ESO combines data obtained with the VIMOS instrument in the U- and R-bands, as well as data obtained in the B-band with the Wide-Field Imager (WFI) attached to the 2.2 m MPG/ESO telescope at La Silla, in the framework of the GABODS survey. The newly released U-band image - the result of 40 hours of staring at the same region of the sky and just made ready by the GOODS team - is the deepest image ever taken from the ground in this wavelength domain. At these depths, the sky is almost completely covered by galaxies, each one, like our own galaxy, the Milky Way, home of hundreds of billions of stars. Galaxies were detected that are a billion times fainter than the unaided eye can see and over a range of colours not directly observable by the eye. This deep image has been essential to the discovery of a large number of new galaxies

  11. Addressing data privacy in matched studies via virtual pooling.

    Science.gov (United States)

    Saha-Chaudhuri, P; Weinberg, C R

    2017-09-07

    Data confidentiality and shared use of research data are two desirable but sometimes conflicting goals in research with multi-center studies and distributed data. While ideal for straightforward analysis, confidentiality restrictions forbid creation of a single dataset that includes covariate information of all participants. Current approaches such as aggregate data sharing, distributed regression, meta-analysis and score-based methods can have important limitations. We propose a novel application of an existing epidemiologic tool, specimen pooling, to enable confidentiality-preserving analysis of data arising from a matched case-control, multi-center design. Instead of pooling specimens prior to assay, we apply the methodology to virtually pool (aggregate) covariates within nodes. Such virtual pooling retains most of the information used in an analysis with individual data and since individual participant data is not shared externally, within-node virtual pooling preserves data confidentiality. We show that aggregated covariate levels can be used in a conditional logistic regression model to estimate individual-level odds ratios of interest. The parameter estimates from the standard conditional logistic regression are compared to the estimates based on a conditional logistic regression model with aggregated data. The parameter estimates are shown to be similar to those without pooling and to have comparable standard errors and confidence interval coverage. Virtual data pooling can be used to maintain confidentiality of data from multi-center study and can be particularly useful in research with large-scale distributed data.

  12. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  13. Study on velocity distribution in a pool by submersible mixers

    International Nuclear Information System (INIS)

    Tian, F; Shi, W D; Lu, X N; Chen, B; Jiang, H

    2012-01-01

    To study the distribution of submersible mixers and agitating effect in the sewage treatment pool, Pro/E software was utilized to build the three-dimensional model. Then, the large-scale computational fluid dynamics software FLUENT6.3 was used. ICEM software was used to build unstructured grid of sewage treatment pool. After that, the sewage treatment pool was numerically simulated by dynamic coordinate system technology and RNG k-ε turbulent model and PIOS algorithm. The macro fluid field and each section velocity flow field distribution were analyzed to observe the efficiency of each submersible mixer. The average velocity and mixing area in the sewage pool were studied simultaneously. Results show that: the preferred project B, two submersible mixers speed is 980 r/min, and setting angles are all 30°. Fluid mixing area in the pool has reached more than 95%. Under the action of two mixers, the fluid in the sewage pool form a continuous circulating water flow. The fluid is mixed adequately and average velocity of fluid in the pool is at around 0.241m/s, which agreed with the work requirements. Consequently it can provide a reference basis for practical engineering application of submersible mixers by using this method.

  14. Absorption process for removing krypton from the off-gas of an LMFBR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Stephenson, M.J.; Dunthorn, D.I.; Reed, W.D.; Pashley, J.H.

    1975-01-01

    The Oak Ridge Gaseous Diffusion Plant selective absorption process for the collection and recovery of krypton and xenon is being further developed to demonstrate, on a pilot scale, a fluorocarbon-based process for removing krypton from the off-gas of an LMFBR fuel reprocessing plant. The new ORGDP selective absorption pilot plant consists of a primary absorption-stripping operation and all peripheral equipment required for feed gas preparation, process solvent recovery, process solvent purification, and krypton product purification. The new plant is designed to achieve krypton decontamination factors in excess of 10 3 with product concentration factors greater than 10 4 while processing a feed gas containing typical quantities of common reprocessing plant off-gas impurities, including oxygen, carbon dioxide, nitrogen oxides, water, xenon, iodine, and methyl iodide. Installation and shakedown of the facility were completed and some short-term tests were conducted early this year. The first operating campaign using a simulated reprocessing plant off-gas feed is now underway. The current program objective is to demonstrate continuous process operability and performance for extended periods of time while processing the simulated ''dirty'' feed. This year's activity will be devoted to routine off-gas processing with little or no deliberate system perturbations. Future work will involve the study of the system behavior under feed perturbations and various plant disturbances. (U.S.)

  15. Fluid structure interaction in LMFBR cores modelling by an homogenization method

    International Nuclear Information System (INIS)

    Brochard, D.

    1988-01-01

    The upper plenum of the internals of PWR, the steam generator bundle, the nuclear reactor core, may be schematically represented by a beam bundle immersed in a fluid. The dynamical study of such a system needs to take into account fluid structure interaction. A refined model at the scale of the tubes can be used but leads to a very difficult problem to solve even on the largest computers. The homogenization method allows to have an approximation of the fluid structure interaction for the global behaviour of the bundle. It consists of replacing the heterogeneous physical medium (tubes and fluid) by an equivalent homogeneous medium whose characteristics are determined from the resolution of a set of problems on the elementary cell. The aim of this paper is to present the main steps of the determination of this equivalent medium in the case of small displacements (acoustic behaviour of the fluid). Then an application to LMFBR core geometry has been realised, which shows the lowering effect on eigenfrequencies due to the fluid. Some comparisons with test results will be presented. 6 refs, 7 figs, 2 tabs

  16. Development of computer code models for analysis of subassembly voiding in the LMFBR

    International Nuclear Information System (INIS)

    Hinkle, W.

    1979-12-01

    The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered

  17. Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

    International Nuclear Information System (INIS)

    Langhans, J.

    1991-01-01

    Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

  18. Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1979-12-01

    The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made.

  19. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  20. Studies needed to prevent the use of expansion bends in LMFBR intermediate heat exchangers

    International Nuclear Information System (INIS)

    Kayser, G.

    1975-01-01

    The LMFBR IHX built in France consist in a vertical tube bundle welded on 2 tube sheets. The secondary sodium flows down a central pipe to the lower collector then up through the tube bundle where it is heated. The solution of the problems raised by the presence of thermal stresses needed thorough studies and led to the following theoretical and experimental developments: 1. A computer code was written for structural analysis. The structure was divided in annular elements that could be studied by means of the elementary theory of shells and plates; and reduced elastic coefficients were given to the tube sheets to account for the presence of drilled holes. 2. An experimental study was undertaken to determine the reduced elastic coefficients of the tube sheets. 3. A computer code was written to study the primary sodium flow around the tube bundle, and experimental studies were made on a mockup, the fluid being water. 4. The results of the previous code were used to determine, by means of a code for thermal analysis, the temperature field in the bundle both in steady state and transient regimes. Up to now, many transients were performed and the Phenix heat exchangers have been operating quite satisfactorily; this seems to prove the design assumptions were correct. (Auth.)