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

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

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)

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

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

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

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

Thermal ionization and plasma state of high temperature vapor of UO2, Cs, and Na: Effect on the heat and radiation transport properties of the vapor phase

International Nuclear Information System (INIS)

Karow, H.U.

1979-01-01

The paper deals with the question how far the thermophysical state and the convective and radiative heat transport properties of vaporized reactor core materials are affected by the thermal ionization existing in the actual vapor state. The materials under consideration here are: nuclear oxide fuel (UO 2 ), Na (as the LMFBR coolant material), and Cs (alkaline fission product, partly retained in the fuel of the core zone). (orig./RW) [de

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

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

Flow visualization in heat-generating porous media

International Nuclear Information System (INIS)

Lee, D.O.; Nilson, R.H.

1977-11-01

The work reported is in support of the Sandia Post-Accident Heat Removal Program, in which simulated LMFBR beds will be subjected to in-pile heating in the ACPR (Annular Core Pulsed Reactor). Flow visualization experiments were performed to gain some insight into the flow patterns and temperature distributions in a fluid-saturated heat-generating porous medium. Although much of the information presented is of a qualitative nature, it is useful in the recognition of the controlling transport process and in the formulation of analytic and numerical models

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)

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

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.)

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.)

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.)

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

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

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

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

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

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)

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

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

KANDY - a numerical model to describe phenomena, which - in a heated and voided fuel element of an LMFBR - may occur

International Nuclear Information System (INIS)

Thurnay, K.

1984-02-01

Kandy is a model developed to describe the essential destructionphenomena of the fuel elements of an LMFBR. The fuel element is assumed to be a voided one, in which the heat generation is still going on. The main process to be modeled is the melting/bursting/evaporating of parts of the fuel pins and the subsequent dislocation of these materials in the coolant channel. The work presented summarizes the assumptions constituting the model, develops the corresponding equations of motion and describes the procedure, turning these into a system of difference-equations ready for coding. As a final part results of a testcase calculation with the Kandy-code are presentend and interpreted. (orig.) [de

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

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

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

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.)

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

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

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)

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

Study on decay heat removal capability of reactor vessel auxiliary cooling system

International Nuclear Information System (INIS)

Nishi, Y.; Kinoshita, I.

1991-01-01

The reactor vessel auxiliary cooling system (RVACS) is a simple, Passive decay heat removal system for an LMFBR. However, the heat removal capacity of this system is small compared to that of an immersed type of decay heat exchanger. In this study, a high-porosity porous body is proposed to enhance the RVACS's heat transfer performance to improve its applicability. The objectives of this study are to propose a new method which is able to use thermal radiation effectively, to confirm its heat removal capability and to estimate its applicability limit of RVACS for an LMFBR. Heat transfer tests were conducted in an experimental facility with a 3.5 m heat transfer height to evaluate the heat transfer performance of the high-porosity porous body. Using the experimental results, plant transient analyses were performed for a 300 MWe pool type LMFBR under a Total Black Out (TBO) condition to confirm the heat removal capability. Furthermore, the relationship between heat removal capability and thermal output of a reactor were evaluated using a simple parameter model

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.

Paleoclassical electron heat transport

International Nuclear Information System (INIS)

Callen, J.D.

2005-01-01

Radial electron heat transport in low collisionality, magnetically-confined toroidal plasmas is shown to result from paleoclassical Coulomb collision processes (parallel electron heat conduction and magnetic field diffusion). In such plasmas the electron temperature equilibrates along magnetic field lines a long length L, which is the minimum of the electron collision length and a maximum effective half length of helical field lines. Thus, the diffusing field lines induce a radial electron heat diffusivity M ≅ L/(πR 0q ) ∼ 10 >> 1 times the magnetic field diffusivity η/μ 0 ≅ ν e (c/ω p ) 2 . The paleoclassical electron heat flux model provides interpretations for many features of 'anomalous' electron heat transport: magnitude and radial profile of electron heat diffusivity (in tokamaks, STs, and RFPs), Alcator scaling in high density plasmas, transport barriers around low order rational surfaces and near a separatrix, and a natural heat pinch (or minimum temperature gradient) heat flux form. (author)

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)

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)

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

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

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

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)

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

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

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

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.)

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)

Heat transfer--Orlando (Symposium), 1980

International Nuclear Information System (INIS)

Stein, R.P.

1980-01-01

This conference proceedings contains 36 papers of which 3 appear as abstracts. 23 papers are indexed separately. Topics covered include: thermodynamics of PWR and LMFBR Steam Generators; two-phase flow in parallel channels; geothermal heat transfer; natural circulation in complex geometries; heat transfer in non-Newtonian systems; and process heat transfer

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

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

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.)

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)

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

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

Evaluation of organic coolants for the transportation of LMFBR spent fuel rods

International Nuclear Information System (INIS)

Arnold, C. Jr.

1978-05-01

The physical and chemical processes that are likely to occur when sodium coated LMFBR spent fuel rods are submerged in various aromatic organic coolants was defined by means of immersion experiments carried out with sodium coated 304 stainless steel coupons. Upon immersion of sodium coated coupons at 220 0 C in hydrocarbon type coolants such as Therminol 88, a mixture of terphenyls, not only was the metallic sodium retained on the coupon, but a carbonaceous coating formed on the surface of the sodium. In contrast, coolants that contained aromatic ether bonds, such as Dowtherm A, reacted with sodium at 220 0 C to form phenolate and other salts, which precipitated from the coolant in the form of a dark sludge. With Dowtherm A, removal of metallic sodium from the coupon was essentially complete in a matter of hours at temperatures of 160--220 0 C. Data on the rate and efficiency of sodium removal upon immersion in Dowtherm A at elevated temperatures were obtained. In addition the kinetics and chemistry of the sodium/Dowtherm A reaction were defined. Because sodium sludges are potentially incompatible with the containing structural materials and the fuel elements, it is recommended that sodium be removed prior to immersion in the coolant via reaction with benzoic acid; this method should be adaptable to the facilities at reactor sites. In aging studies Dowtherm A was found to be thermally stable up to 400 0 C and radiatively stable at ambient conditions. The combined effect of heat and radiation was not defined

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

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)

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)

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

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

Heat transport and storage

International Nuclear Information System (INIS)

Despois, J.

1977-01-01

Recalling the close connections existing between heat transport and storage, some general considerations on the problem of heat distribution and transport are presented 'in order to set out the problem' of storage in concrete form. This problem is considered in its overall plane, then studied under the angle of the different technical choices it involves. The two alternatives currently in consideration are described i.e.: storage in a mined cavity and underground storage as captive sheet [fr

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

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

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.

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

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

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

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

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.)

Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

International Nuclear Information System (INIS)

1980-01-01

Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

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)

Method for pre-heating lmfbr type reactors

International Nuclear Information System (INIS)

Yokozawa, Atsushi; Kataoka, Hajime.

1978-01-01

Purpose: To enable pre-heating for the inside of the reactor container and the inside of the coolant recycling system with no additional facilities. Method: The coolant recycling system is composed of a heat exchanger, a mechanical pump, a check valve, a flow meter or the like and it is connected in series by way of a pipe line to a reactor container. The mechanical pump is used as a gas recycling device upon pre-heating and it is designed so that a blower such as a fan can be replaced for the impeller of the pump. The inside of the reactor container and the inside of the coolant recycling system is at first filled with an inert gas such as for use with cover gas. Then, nuclear fuels are loaded to attain criticality. Simultaneously, the blower is started and the control rods are operated while cooling the nuclear fuel with the inert gas thus to obtain heat required for pre-heating the pipe line or the like from the nuclear fuels. After the completion of the pre-heating, the liquid metal is charged. (Ikeda, J.)

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.)

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

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

SEAWAT-based simulation of axisymmetric heat transport.

Science.gov (United States)

Vandenbohede, Alexander; Louwyck, Andy; Vlamynck, Nele

2014-01-01

Simulation of heat transport has its applications in geothermal exploitation of aquifers and the analysis of temperature dependent chemical reactions. Under homogeneous conditions and in the absence of a regional hydraulic gradient, groundwater flow and heat transport from or to a well exhibit radial symmetry, and governing equations are reduced by one dimension (1D) which increases computational efficiency importantly. Solute transport codes can simulate heat transport and input parameters may be modified such that the Cartesian geometry can handle radial flow. In this article, SEAWAT is evaluated as simulator for heat transport under radial flow conditions. The 1971, 1D analytical solution of Gelhar and Collins is used to compare axisymmetric transport with retardation (i.e., as a result of thermal equilibrium between fluid and solid) and a large diffusion (conduction). It is shown that an axisymmetric simulation compares well with a fully three dimensional (3D) simulation of an aquifer thermal energy storage systems. The influence of grid discretization, solver parameters, and advection solution is illustrated. Because of the high diffusion to simulate conduction, convergence criterion for heat transport must be set much smaller (10(-10) ) than for solute transport (10(-6) ). Grid discretization should be considered carefully, in particular the subdivision of the screen interval. On the other hand, different methods to calculate the pumping or injection rate distribution over different nodes of a multilayer well lead to small differences only. © 2013, National Ground Water Association.

A simple Boltzmann transport equation for ballistic to diffusive transient heat transport

International Nuclear Information System (INIS)

Maassen, Jesse; Lundstrom, Mark

2015-01-01

Developing simplified, but accurate, theoretical approaches to treat heat transport on all length and time scales is needed to further enable scientific insight and technology innovation. Using a simplified form of the Boltzmann transport equation (BTE), originally developed for electron transport, we demonstrate how ballistic phonon effects and finite-velocity propagation are easily and naturally captured. We show how this approach compares well to the phonon BTE, and readily handles a full phonon dispersion and energy-dependent mean-free-path. This study of transient heat transport shows (i) how fundamental temperature jumps at the contacts depend simply on the ballistic thermal resistance, (ii) that phonon transport at early times approach the ballistic limit in samples of any length, and (iii) perceived reductions in heat conduction, when ballistic effects are present, originate from reductions in temperature gradient. Importantly, this framework can be recast exactly as the Cattaneo and hyperbolic heat equations, and we discuss how the key to capturing ballistic heat effects is to use the correct physical boundary conditions

Analysis of transient thermal response in the outlet plenum of an LMFBR

International Nuclear Information System (INIS)

Yang, J.W.

1976-05-01

A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter 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 bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC

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

One-Loop Operation of Primary Heat Transport System in MONJU During Heat Transport System Modifications

International Nuclear Information System (INIS)

Goto, T.; Tsushima, H.; Sakurai, N.; Jo, T.

2006-01-01

MONJU is a prototype fast breeder reactor (FBR). Modification work commenced in March 2005. Since June 2004, MONJU has changed to one-loop operation of the primary heat transport system (PHTS) with all of the secondary heat transport systems (SHTS) drained of sodium. The purposes of this change are to shorten the modification period and to reduce the cost incurred for circuit trace heating electrical consumption. Before changing condition, the following issues were investigated to show that this mode of operation was possible. The heat loss from the reactor vessel and the single primary loop must exceed the decay heat by an acceptable margin but the capacity of pre-heaters to keep the sodium within the primary vessel at about 200 deg. C must be maintained. With regard to the heat loss and the decay heat, the estimated heat loss in the primary system was in the range of 90-170 kW in one-loop operation, and the calculated decay heat was 21.2 kW. Although the heat input of the primary pump was considered, it was clear that circuit heat loss greatly exceeded the decay heat. As for pre-heaters, effective capacity was less than the heat loss. Therefore, the temperature of the reactor vessel room was raised to reduce the heat loss. One-loop operation of the PHTS was able to be executed by means of these measures. The cost of electrical consumption in the power plant has been reduced by one-loop operation of the PHTS and the modification period was shortened. (authors)

Simulation of heat and mass transfer processes in molten core debris-concrete systems. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Felde, D K

1979-01-01

The heat and mass transport phenomena taking place in volumetrically-heated fluids have become of interest in recent years due to their significance in assessments of fast reactor safety and post-accident heat removal (PAHR). Following a hypothetical core disruptive accident (HCDA), the core and reactor internals may melt down. The core debis melting through the reactor vessel and guard vessel may eventually contact the concrete of the reactor cell floor. The interaction of the core debris with the concrete as well as the melting of the debris pool into the concrete will significantly affect efforts to prevent breaching of the containment and the resultant release of radioactive effluents to the environment.

Particle and heat transport in Tokamaks

International Nuclear Information System (INIS)

Chatelier, M.

1984-01-01

A limitation to performances of tokamaks is heat transport through magnetic surfaces. Principles of ''classical'' or ''neoclassical'' transport -i.e. transport due to particle and heat fluxes due to Coulomb scattering of charged particle in a magnetic field- are exposed. It is shown that beside this classical effect, ''anomalous'' transport occurs; it is associated to the existence of fluctuating electric or magnetic fields which can appear in the plasma as a result of charge and current perturbations. Tearing modes and drift wave instabilities are taken as typical examples. Experimental features are presented which show that ions behave approximately in a classical way whereas electrons are strongly anomalous [fr

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

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

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)

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)

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.)

Heat in the Barents Sea: transport, storage, and surface fluxes

Directory of Open Access Journals (Sweden)

L. H. Smedsrud

2010-02-01

Full Text Available A column model is set up for the Barents Sea to explore sensitivity of surface fluxes and heat storage from varying ocean heat transport. Mean monthly ocean transport and atmospheric forcing are synthesised and force the simulations. Results show that by using updated ocean transports of heat and freshwater the vertical mean hydrographic seasonal cycle can be reproduced fairly well.

Our results indicate that the ~70 TW of heat transported to the Barents Sea by ocean currents is lost in the southern Barents Sea as latent, sensible, and long wave radiation, each contributing 23–39 TW to the total heat loss. Solar radiation adds 26 TW in the south, as there is no significant ice production.

The northern Barents Sea receives little ocean heat transport. This leads to a mixed layer at the freezing point during winter and significant ice production. There is little net surface heat loss annually in the north. The balance is achieved by a heat loss through long wave radiation all year, removing most of the summer solar heating.

During the last decade the Barents Sea has experienced an atmospheric warming and an increased ocean heat transport. The Barents Sea responds to such large changes by adjusting temperature and heat loss. Decreasing the ocean heat transport below 50 TW starts a transition towards Arctic conditions. The heat loss in the Barents Sea depend on the effective area for cooling, and an increased heat transport leads to a spreading of warm water further north.

Heat transport and surface heat transfer with helium in rotating channels

International Nuclear Information System (INIS)

Schnapper, C.

1978-06-01

Heat transport and surface heat transfer with helium in rotating radially arranged channels were experimentally studied with regard to cooling of large turbogenerators with superconducting windings. Measurements with thermosiphon and thermosiphon loops of different channel diameters were performed, and results are presented. The thermodynamic state of the helium in a rotating thermosiphon and the mass flow rate in a thermosiphon loop is characterized by formulas. Heat transport by directed convection in thermosiphon loops is found to be more efficient 12 cm internal convection in thermosiphons. Steady state is reached sooner in thermosiphon loops than in thermosiphons, when heat load suddenly changes. In a very large centrifugal field single-phase heat transfer with natural and forced convection is described by similar formulas which are also applicable 10 thermosiphons in gravitation field or to heat transfer to non-rotating helium. (orig.) [de

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

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)

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)

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)

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)

Understanding of flux-limited behaviors of heat transport in nonlinear regime

Energy Technology Data Exchange (ETDEWEB)

Guo, Yangyu, E-mail: yangyuhguo@gmail.com [Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Engineering Mechanics and CNMM, Tsinghua University, Beijing 100084 (China); Jou, David, E-mail: david.jou@uab.es [Departament de Física, Universitat Autònoma de Barcelona, 08193 Bellaterra, Catalonia (Spain); Wang, Moran, E-mail: mrwang@tsinghua.edu [Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Engineering Mechanics and CNMM, Tsinghua University, Beijing 100084 (China)

2016-01-28

The classical Fourier's law of heat transport breaks down in highly nonequilibrium situations as in nanoscale heat transport, where nonlinear effects become important. The present work is aimed at exploring the flux-limited behaviors based on a categorization of existing nonlinear heat transport models in terms of their theoretical foundations. Different saturation heat fluxes are obtained, whereas the same qualitative variation trend of heat flux versus exerted temperature gradient is got in diverse nonlinear models. The phonon hydrodynamic model is proposed to act as a standard to evaluate other heat flux limiters because of its more rigorous physical foundation. A deeper knowledge is thus achieved about the phenomenological generalized heat transport models. The present work provides deeper understanding and accurate modeling of nonlocal and nonlinear heat transport beyond the diffusive limit. - Highlights: • Exploring flux-limited behaviors based on a categorization of existing nonlinear heat transport models. • Proposing phonon hydrodynamic model as a standard to evaluate heat flux limiters. • Providing accurate modeling of nonlocal and nonlinear heat transport beyond the diffusive limit.

Theory of ion heat transport in tokamaks

International Nuclear Information System (INIS)

Gott, Y.V.; Yurchenko, E.I.

1987-01-01

Experiments which have been carried out in several tokamaks to determine the ion thermal conductivity show that it is several times the value predicted by the neoclassical theory. A possible explanation for this discrepancy is proposed. When the finite width of a banana is taken into account, there are substantial increases in the heat fluxes which stem from the important contribution of superthermal ions to the transport. If the electron diffusive flux is zero, a systematic account of the ions with E>T leads to an ion heat flux with a finite banana width which is two to four times the neoclassical prediction. The effect of the anomalous nature of the electron flux on the ion heat transport is analyzed. An expression is derived for calculating the ion heat transport over the entire range of collision rates

Progress in understanding heat transport at JET

International Nuclear Information System (INIS)

Mantica, P.; Garbet, X.; Angioni, C.

2005-01-01

This paper reports recent progress in understanding heat transport mechanisms either in conventional or advanced tokamak scenarios in JET. A key experimental tool has been the use of perturbative transport techniques, both by ICH power modulation and by edge cold pulses. The availability of such results has allowed careful comparison with theoretical modelling using 1D empirical or physics based transport models, 3D fluid turbulence simulations or gyrokinetic stability analysis. In conventional L- and H-mode plasmas the issue of temperature profile stiffness has been addressed. JET results are consistent with the concept of a critical inverse temperature gradient length above which transport is enhanced by the onset of turbulence. A threshold value R/L Te ∼5 has been found for the onset of stiff electron transport, while the level of electron stiffness appears to vary strongly with plasma parameters, in particular with the ratio of electron and ion heating: electrons become stiffer when ions are strongly heated, resulting in larger R/L Ti values. This behaviour has also been found theoretically, although quantitatively weaker than in experiments. In plasmas characterized by Internal Transport Barriers (ITB), the properties of heat transport inside the ITB layer and the ITB formation mechanisms have been investigated. The plasma current profile is found to play a major role in ITB formation. The effect of negative magnetic shear on electron and ion stabilization is demonstrated both experimentally and theoretically using turbulence codes. The role of rational magnetic surfaces in ITB triggering is well assessed experimentally, but still lacks a convincing theoretical explanation. Attempts to trigger an ITB by externally induced magnetic reconnection using saddle coils have shown that MHD islands in general do not produce a sufficient variation of ExB flow shear to lead to ITB formation. First results of perturbative transport in ITBs show that the ITB is a narrow

Active transport and heat.

Science.gov (United States)

Tait, Peter W

2011-07-01

Increasing heat may impede peoples' ability to be active outdoors thus limiting active transport options. Co-benefits from mitigation of and adaptation to global warming should not be assumed but need to be actively designed into strategies.

Possible role of oceanic heat transport in early Eocene climate

Science.gov (United States)

Sloan, L. C.; Walker, J. C.; Moore, T. C. Jr

1995-01-01

Increased oceanic heat transport has often been cited as a means of maintaining warm high-latitude surface temperatures in many intervals of the geologic past, including the early Eocene. Although the excess amount of oceanic heat transport required by warm high latitude sea surface temperatures can be calculated empirically, determining how additional oceanic heat transport would take place has yet to be accomplished. That the mechanisms of enhanced poleward oceanic heat transport remain undefined in paleoclimate reconstructions is an important point that is often overlooked. Using early Eocene climate as an example, we consider various ways to produce enhanced poleward heat transport and latitudinal energy redistribution of the sign and magnitude required by interpreted early Eocene conditions. Our interpolation of early Eocene paleotemperature data indicate that an approximately 30% increase in poleward heat transport would be required to maintain Eocene high-latitude temperatures. This increased heat transport appears difficult to accomplish by any means of ocean circulation if we use present ocean circulation characteristics to evaluate early Eocene rates. Either oceanic processes were very different from those of the present to produce the early Eocene climate conditions or oceanic heat transport was not the primary cause of that climate. We believe that atmospheric processes, with contributions from other factors, such as clouds, were the most likely primary cause of early Eocene climate.

The experiment study of the thermal insulation of the roof-slab of the main vessel of a LMFBR

International Nuclear Information System (INIS)

Wang Zhifeng; Wang Zhou; Yang Xianyong

1995-01-01

The effects of composition of insulation, i.e., reflective multi-plate thermal insulator, protecting the roof-slab of the vessel of the LMFBR on the heat transfer performance has been studied experimentally for CEFR. A economical form of the thermal insulation is suggested for CEFR. In addition, the scheme without reflective thermal insulator which has only a forced convection cooling system has been studied experimentally and a formula to calculate the average Nusselt number of the flow channel, which is valuable for CEFR design, has been raised

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

Coupled heat transfer in high temperature transporting system with semitransparent/opaque material

International Nuclear Information System (INIS)

Du Shenghua; Xia Xinjin

2010-01-01

The heat transfer model of the aerodynamic heating coupled with radiative cooling was developed. The thermal protect system includes the higher heat flux region with high temperature semitransparent material, the heat transporting channel and the lower heat flux region with metal. The control volume method was combined with the Monte Carlo method to calculate the coupled heat transfer of the transporting system, and the thermal equilibrium equation for the transporting channel was solved simultaneously. The effect of the aeroheating flux radio, the area ratio of radiative surfaces, the convective heat transfer coefficient of the heat transporting channel on the radiative surface temperature and the fluid temperature in the heat transporting channel were analyzed. The effect of radiation and conduction in the semitransparent material was discussed. The result shows that to increase the convective heat transfer coefficient in heat flux channel can enhance the heat transporting ability of the system, but the main parameter to effect on the temperature of the heat transporting system is the area ratio of radiative surfaces. (authors)

Heat Transfer in Directional Water Transport Fabrics

Directory of Open Access Journals (Sweden)

Chao Zeng

2016-10-01

Full Text Available Directional water transport fabrics can proactively transfer moisture from the body. They show great potential in making sportswear and summer clothing. While moisture transfer has been previously reported, heat transfer in directional water transport fabrics has been little reported in research literature. In this study, a directional water transport fabric was prepared using an electrospraying technique and its heat transfer properties under dry and wet states were evaluated, and compared with untreated control fabric and the one pre-treated with NaOH. All the fabric samples showed similar heat transfer features in the dry state, and the equilibrium temperature in the dry state was higher than for the wet state. Wetting considerably enhanced the thermal conductivity of the fabrics. Our studies indicate that directional water transport treatment assists in moving water toward one side of the fabric, but has little effect on thermal transfer performance. This study may be useful for development of “smart” textiles for various applications.

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

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

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

Ion heat transport studies in JET

DEFF Research Database (Denmark)

Mantica, P; Angioni, C; Baiocchi, B

2011-01-01

Detailed experimental studies of ion heat transport have been carried out in JET exploiting the upgrade of active charge exchange spectroscopy and the availability of multi-frequency ion cyclotron resonance heating with 3He minority. The determination of ion temperature gradient (ITG) threshold a...

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

Electron and ion heat transport with lower hybrid current drive and neutral beam injection heating in ASDEX

International Nuclear Information System (INIS)

Soeldner, F.X.; Pereverzev, G.V.; Bartiromo, R.; Fahrbach, H.U.; Leuterer, F.; Murmann, H.D.; Staebler, A.; Steuer, K.H.

1993-01-01

Transport code calculations were made for experiments with the combined operation of lower hybrid current drive and heating and of neutral beam injection heating on ASDEX. Peaking or flattening of the electron temperature profile are mainly explained by modifications of the MHD induced electron heat transport. They originate from current profile changes due to lower hybrid and neutral beam current drive and to contributions from the bootstrap current. Ion heat transport cannot be described by one single model for all heating scenarios. The ion heat conductivity is reduced during lower hybrid heated phases with respect to Ohmic and neutral beam heating. (author). 13 refs, 5 figs

An Overview of Liquid Fluoride Salt Heat Transport Systems

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL

2010-09-01

Heat transport is central to all thermal-based forms of electricity generation. The ever increasing demand for higher thermal efficiency necessitates power generation cycles transitioning to progressively higher temperatures. Similarly, the desire to provide direct thermal coupling between heat sources and higher temperature chemical processes provides the underlying incentive to move toward higher temperature heat transfer loops. As the system temperature rises, the available materials and technology choices become progressively more limited. Superficially, fluoride salts at {approx}700 C resemble water at room temperature being optically transparent and having similar heat capacity, roughly three times the viscosity, and about twice the density. Fluoride salts are a leading candidate heat-transport material at high temperatures. Fluoride salts have been extensively used in specialized industrial processes for decades, yet they have not entered widespread deployment for general heat transport purposes. This report does not provide an exhaustive screening of potential heat transfer media and other high temperature liquids such as alkali metal carbonate eutectics or chloride salts may have economic or technological advantages. A particular advantage of fluoride salts is that the technology for their use is relatively mature as they were extensively studied during the 1940s-1970s as part of the U.S. Atomic Energy Commission's program to develop molten salt reactors (MSRs). However, the instrumentation, components, and practices for use of fluoride salts are not yet developed sufficiently for commercial implementation. This report provides an overview of the current understanding of the technologies involved in liquid salt heat transport (LSHT) along with providing references to the more detailed primary information resources. Much of the information presented here derives from the earlier MSR program. However, technology has evolved over the intervening years

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)

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)

Some factors affecting radiative heat transport in PWR cores

International Nuclear Information System (INIS)

Hall, A.N.

1989-04-01

This report discusses radiative heat transport in Pressurized Water Reactor cores, using simple models to illustrate basic features of the transport process. Heat transport by conduction and convection is ignored in order to focus attention on the restrictions on radiative heat transport imposed by the geometry of the heat emitting and absorbing structures. The importance of the spacing of the emitting and absorbing structures is emphasised. Steady state temperature distributions are found for models of cores which are uniformly heated by fission product decay. In all of the models, a steady state temperature distribution can only be obtained if the central core temperature is in excess of the melting point of UO 2 . It has recently been reported that the MIMAS computer code, which takes into account radiative heat transport, has been used to model the heat-up of the Three Mile Island-2 reactor core, and the computations indicate that the core could not have reached the melting point of UO 2 at any time or any place. We discuss this result in the light of the calculations presented in this paper. It appears that the predicted stabilisation of the core temperatures at ∼ 2200 0 C may be a consequence of the artificially large spacing between the radial rings employed in the MIMAS code, rather than a result of physical significance. (author)

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

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

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

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

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

Source effects on impurity and heat transport in a tokamak

International Nuclear Information System (INIS)

Bennett, R.B.

1980-12-01

A recently developed generalization of neoclassical theory is extended here to study heat flux contributions to impurity transport, as well as the heat fluxes themselves. The theory accounts for the first four source moments, with external drags, which has been studied previously with either fewer moments or restricted to a collisional plasma. Conditions are established for which a momentum source may be used to modify the particle and heat transport. In the course of this work, the particle and heat transport is evaluated for a two species plasma with arbitrary plasma geometry, beta, and collisionality

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

Transport lattice models of heat transport in skin with spatially heterogeneous, temperature-dependent perfusion

Directory of Open Access Journals (Sweden)

Martin Gregory T

2004-11-01

Full Text Available Abstract Background Investigation of bioheat transfer problems requires the evaluation of temporal and spatial distributions of temperature. This class of problems has been traditionally addressed using the Pennes bioheat equation. Transport of heat by conduction, and by temperature-dependent, spatially heterogeneous blood perfusion is modeled here using a transport lattice approach. Methods We represent heat transport processes by using a lattice that represents the Pennes bioheat equation in perfused tissues, and diffusion in nonperfused regions. The three layer skin model has a nonperfused viable epidermis, and deeper regions of dermis and subcutaneous tissue with perfusion that is constant or temperature-dependent. Two cases are considered: (1 surface contact heating and (2 spatially distributed heating. The model is relevant to the prediction of the transient and steady state temperature rise for different methods of power deposition within the skin. Accumulated thermal damage is estimated by using an Arrhenius type rate equation at locations where viable tissue temperature exceeds 42°C. Prediction of spatial temperature distributions is also illustrated with a two-dimensional model of skin created from a histological image. Results The transport lattice approach was validated by comparison with an analytical solution for a slab with homogeneous thermal properties and spatially distributed uniform sink held at constant temperatures at the ends. For typical transcutaneous blood gas sensing conditions the estimated damage is small, even with prolonged skin contact to a 45°C surface. Spatial heterogeneity in skin thermal properties leads to a non-uniform temperature distribution during a 10 GHz electromagnetic field exposure. A realistic two-dimensional model of the skin shows that tissue heterogeneity does not lead to a significant local temperature increase when heated by a hot wire tip. Conclusions The heat transport system model of the

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

Intense radiative heat transport across a nano-scale gap

International Nuclear Information System (INIS)

Budaev, Bair V.; Ghafari, Amin; Bogy, David B.

2016-01-01

In this paper, we analyze the radiative heat transport in layered structures. The analysis is based on our prior description of the spectrum of thermally excited waves in systems with a heat flux. The developed method correctly predicts results for all known special cases for both large and closing gaps. Numerical examples demonstrate the applicability of our approach to the calculation of the radiative heat transport coefficient across various layered structures.

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

Modelling of Temperature Profiles and Transport Scaling in Auxiliary Heated Tokamaks

DEFF Research Database (Denmark)

Callen, J.D.; Christiansen, J.P.; Cordey, J.G.

1987-01-01

time , the heating effectiveness η, and the energy offset W(0). Considering both the temperature profile responses and the global transport scaling, the constant heat pinch or excess temperature gradient model is found to best characterize the present JET data. Finally, new methods are proposed......The temperature profiles produced by various heating profiles are calculated from local heat transport models. The models take the heat flux to be the sum of heat diffusion and a non-diffusive heat flow, consistent with local measurements of heat transport. Two models are developed analytically...... in detail: (i) a heat pinch or excess temperature gradient model with constant coefficients; and (ii) a non-linear heat diffusion coefficient (χ) model. Both models predict weak (lesssim20%) temperature profile responses to physically relevant changes in the heat deposition profile – primarily because...

TOUGH, Unsaturated Groundwater Transport and Heat Transport Simulation

International Nuclear Information System (INIS)

Pruess, K.A.; Cooper, C.; Osnes, J.D.

1992-01-01

1 - Description of program or function: A successor to the TOUGH program, TOUGH2 offers added capabilities and user features, including the flexibility to handle different fluid mixtures (water, water with tracer; water, CO 2 ; water, air; water, air with vapour pressure lowering, and water, hydrogen), facilities for processing of geometric data (computational grids), and an internal version control system to ensure referenceability of code applications. TOUGH (Transport of Unsaturated Groundwater and Heat) is a multi-dimensional numerical model for simulating the coupled transport of water, vapor, air, and heat in porous and fractured media. The program provides options for specifying injection or withdrawal of heat and fluids. Although primarily designed for studies of high-level nuclear waste isolation in partially saturated geological media, it should also be useful for a wider range of problems in heat and moisture transfer, and in the drying of porous materials. For example, geothermal reservoir simulation problems can be handled simply by setting the air mass function equal to zero on input. The TOUGH simulator was developed for problems involving strongly heat-driven flow. To describe these phenomena a multi-phase approach to fluid and heat flow is used, which fully accounts for the movement of gaseous and liquid phases, their transport of latent transitions between liquid and vapor. TOUGH takes account of fluid flow in both liquid and gaseous phases occurring under pressure, viscous, and gravity forces according to Darcy's law. Interference between the phases is represented by means of relative permeability functions. The code handles binary, but not Knudsen, diffusion in the gas phase and capillary and phase absorption effects for the liquid phase. Heat transport occurs by means of conduction with thermal conductivity dependent on water saturation, convection, and binary diffusion, which includes both sensible and latent heat. 2 - Method of solution: All

ECRH and electron heat transport in tokamaks

International Nuclear Information System (INIS)

Zou, X.L.; Giruzzi, G.; Dumont, R.J.

2003-01-01

It has been observed during the ECRH experiments in tokamaks that the shape of the electron temperature profile in stationary regimes is not very sensitive to the ECRH power deposition i.e. the temperature profile remains peaked at the center even though the ECRH power deposition is off-axis. Various models have been invoked for the interpretation of this profile resilience phenomenon: the inward heat pinch, the critical temperature gradient, the Self-Organized Criticality, etc. Except the pinch effect, all of these models need a specific form of the diffusivity in the heat transport equation. In this work, our approach is to solve a simplified time-dependent heat transport equation analytically in cylindrical geometry. The features of this analytical solution are analyzed, in particular the relationship between the temperature profile resilience and the Eigenmode of the physical system with respect to the heat transport phenomenon. Finally, applications of this analytical solution for the determination of the transport coefficient and the polarization of the EC waves are presented. It has been shown that the solution of the simplified transport equation in a finite cylinder is a Fourier-Bessel series. This series represents in fact a decomposition of the heat source in Eigenmode, which are characterized by the Bessel functions of order 0. The physical interpretation of the Eigenmodes is the following: when the heat source is given by a Bessel function of order 0, the temperature profile has exactly the same form as the source at every time. At the beginning of the power injection, the effectiveness of the temperature response is the same for each Eigenmode, and the response in temperature, having the same form as the source, is local. Conversely, in the later phase of the evolution, the effectiveness of the temperature response for each Eigenmode is different: the higher the order, the lower the effectiveness. In this case the response in temperature appears as

The use of dielectric heating in particulate bed dryout experiments

International Nuclear Information System (INIS)

Stevens, G.F.; Willshire, S.J.

1984-09-01

Decay-heated, liquid-saturated debris beds arise in hypothetical severe accidents with LMFBR and PWR, and a large international effort is currently engaged in experimental studies of the cooling limitations of such beds. Dryout is one of the important cooling limitations. Dielectric heating offers a means of closely simulating decay heating in beds of irregular particles, and is under development at AEE Winfrith for application to experimental studies of dryout. This report describes progress to date. (author)

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

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

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

Theory and use of GIRAFFE for analysis of decay characteristics of delayed-neutron precursors in an LMFBR

International Nuclear Information System (INIS)

Gross, K.C.

1980-07-01

The application of the computer code GIRAFFE (General Isotope Release Analysis For Failed Elements) written in FORTRAN IV is described. GIRAFFE was designed to provide parameter estimates of the nonlinear discrete-measurement models that govern the transport and decay of delayed-neutron precursors in a liquid-metal fast breeder reactor (LMFBR). The code has been organized into a set of small, relatively independent and well-defined modules to facilitate modification and maintenance. The program logic, the numerical techniques, and the methods of solution used by the code are presented, and the functions of the MAIN program and of each subroutine are discussed

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

Miniature Heat Transport System for Spacecraft Thermal Control

Science.gov (United States)

Ochterbeck, Jay M.; Ku, Jentung (Technical Monitor)

2002-01-01

Loop heat pipes (LHP) are efficient devices for heat transfer and use the basic principle of a closed evaporation-condensation cycle. The advantage of using a loop heat pipe over other conventional methods is that large quantities of heat can be transported through a small cross-sectional area over a considerable distance with no additional power input to the system. By using LHPs, it seems possible to meet the growing demand for high-power cooling devices. Although they are somewhat similar to conventional heat pipes, LHPs have a whole set of unique properties, such as low pressure drops and flexible lines between condenser and evaporator, that make them rather promising. LHPs are capable of providing a means of transporting heat over long distances with no input power other than the heat being transported because of the specially designed evaporator and the separation of liquid and vapor lines. For LHP design and fabrication, preliminary analysis on the basis of dimensionless criteria is necessary because of certain complicated phenomena that take place in the heat pipe. Modeling the performance of the LHP and miniaturizing its size are tasks and objectives of current research. In the course of h s work, the LHP and its components, including the evaporator (the most critical and complex part of the LHP), were modeled with the corresponding dimensionless groups also being investigated. Next, analysis of heat and mass transfer processes in the LHP, selection of the most weighted criteria from known dimensionless groups (thermal-fluid sciences), heat transfer rate limits, (heat pipe theory), and experimental ratios which are unique to a given heat pipe class are discussed. In the third part of the report, two-phase flow heat and mass transfer performances inside the LHP condenser are analyzed and calculated for Earth-normal gravity and microgravity conditions. On the basis of recent models and experimental databanks, an analysis for condensing two-phase flow regimes

Policies and initiatives for carbon neutrality in nordic heating and transport systems

DEFF Research Database (Denmark)

Muller, Jakob Glarbo; Wu, Qiuwei; Ostergaard, Jacob

2012-01-01

Policies and initiatives promoting carbon neutrality in the Nordic heating and transport systems are presented. The focus within heating systems is the propagation of heat pumps while the focus within transport systems is initiatives regarding electric vehicles (EVs). It is found that conversion...... to heat pumps in the Nordic region rely on both private economic and national economic incentives. Initiatives toward carbon neutrality in the transport system are mostly concentrated on research, development and demonstration for deployment of a large number of EVs. All Nordic countries have plans...... for the future heating and transport systems with the ambition of realizing carbon neutrality....

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

Transport in Auxiliary Heated NSTX Discharges

International Nuclear Information System (INIS)

LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitte, M.L.; Bourdelle, C.; Gates, D.A.; Kaye, S.M.; Maingi, R.; Menard, J.E.; Mueller, D.; Ono, M.; Paul, S.F.; Redi, M.H.; Roquemore, A.L.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Synakowski, E.J.; Soukhanovskii, V.A.; Wilson, J.R.

2003-01-01

The NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses

Investigation of the physical and numerical foundations of two-fluid representation of sodium boiling with applications to LMFBR experiments

International Nuclear Information System (INIS)

No, H.C.; Kazimi, M.S.

1983-03-01

This work involves the development of physical models for the constitutive relations of a two-fluid, three-dimensional sodium boiling code, THERMIT-6S. The code is equipped with a fluid conduction model, a fuel pin model, and a subassembly wall model suitable for stimulating LMFBR transient events. Mathematically rigorous derivations of time-volume averaged conservation equations are used to establish the differential equations of THERMIT-6S. These equations are then discretized in a manner identical to the original THERMIT code. A virtual mass term is incorporated in THERMIT-6S to solve the ill-posed problem. Based on a simplified flow regime, namely cocurrent annular flow, constitutive relations for two-phase flow of sodium are derived. The wall heat transfer coefficient is based on momentum-heat transfer analogy and a logarithmic law for liquid film velocity distribution. A broad literature review is given for two-phase friction factors. It is concluded that entrainment can account for some of the discrepancies in the literature. Mass and energy exchanges are modelled by generalization of the turbulent flux concept. Interfacial drag coefficients are derived for annular flows with entrainment. Code assessment is performed by simulating three experiments for low flow-high power accidents and one experiment for low flow/low power accidents in the LMFBR. While the numerical results for pre-dryout are in good agreement with the data, those for post-dryout reveal the need for improvement of the physical models. The benefits of two-dimensional non-equilibrium representation of sodium boiling are studied

Electron heat transport analysis of low-collisionality plasmas in the neoclassical-transport-optimized configuration of LHD

International Nuclear Information System (INIS)

Murakami, Sadayoshi; Yamada, Hiroshi; Wakasa, Arimitsu

2002-01-01

Electron heat transport in low-collisionality LHD plasma is investigated in order to study the neoclassical transport optimization effect on thermal plasma transport with an optimization level typical of so-called ''advanced stellarators''. In the central region, a higher electron temperature is obtained in the optimized configuration, and transport analysis suggests the considerable effect of neoclassical transport on the electron heat transport assuming the ion-root level of radial electric field. The obtained experimental results support future reactor design in which the neoclassical and/or anomalous transports are reduced by magnetic field optimization in a non-axisymmetric configuration. (author)

Calculation of Doses Due to Accidentally Released Plutonium From An LMFBR

Energy Technology Data Exchange (ETDEWEB)

Fish, B.R.

2001-08-07

Experimental data and analytical models that should be considered in assessing the transport properties of plutonium aerosols following a hypothetical reactor accident have been examined. Behaviors of released airborne materials within the reactor containment systems, as well as in the atmosphere near the reactor site boundaries, have been semiquantitatively predicted from experimental data and analytical models. The fundamental chemistry of plutonium as it may be applied in biological systems has been used to prepare models related to the intake and metabolism of plutonium dioxide, the fuel material of interest. Attempts have been made to calculate the possible doses from plutonium aerosols for a typical analyzed release in order to evaluate the magnitude of the internal exposure hazards that might exist in the vicinity of the reactor after a hypothetical LMFBR (Liquid-Metal Fast Breeder Reactor) accident. Intake of plutonium (using data for {sup 239}Pu as an example) and its distribution in the body were treated parametrically without regard to the details of transport pathways in the environment. To the extent possible, dose-response data and models have been reviewed, and an assessment of their adequacy has been made so that recommended or preferred practices could be developed.

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

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

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

Generalized heat-transport equations: parabolic and hyperbolic models

Science.gov (United States)

Rogolino, Patrizia; Kovács, Robert; Ván, Peter; Cimmelli, Vito Antonio

2018-03-01

We derive two different generalized heat-transport equations: the most general one, of the first order in time and second order in space, encompasses some well-known heat equations and describes the hyperbolic regime in the absence of nonlocal effects. Another, less general, of the second order in time and fourth order in space, is able to describe hyperbolic heat conduction also in the presence of nonlocal effects. We investigate the thermodynamic compatibility of both models by applying some generalizations of the classical Liu and Coleman-Noll procedures. In both cases, constitutive equations for the entropy and for the entropy flux are obtained. For the second model, we consider a heat-transport equation which includes nonlocal terms and study the resulting set of balance laws, proving that the corresponding thermal perturbations propagate with finite speed.

Results from transient transport experiments in Rijnhuizen tokamak project: Heat convection, transport barriers and 'non-local' effects

International Nuclear Information System (INIS)

Mantica, P.; Gorini, G.; Hogeweij, G.M.D.; Kloe, J. de; Lopez Cardozo, N.J.; Schilham, A.M.R.

2001-01-01

An overview of experimental transport studies performed on the Rijnhuizen Tokamak Project (RTP) using transient transport techniques in both Ohmic and ECH dominated plasmas is presented. Modulated Electron Cyclotron Heating (ECH) and oblique pellet injection (OPI) have been used to induce electron temperature (T e ) perturbations at different radial locations. These were used to probe the electron transport barriers observed near low order rational magnetic surfaces in ECH dominated steady-state RTP plasmas. Layers of inward electron heat convection in off-axis ECH plasmas were detected with modulated ECH. This suggests that RTP electron transport barriers consist of heat pinch layers rather than layers of low thermal diffusivity. In a different set of experiments, OPI triggered a transient rise of the core T e due to an increase of the T e gradient in the 1< q<2 region. These transient transport barriers were probed with modulated ECH and found to be due to a transient drop of the electron heat diffusivity, except for off-axis ECH plasmas, where a transient inward pinch is also observed. Transient transport studies in RTP could not solve this puzzling interplay between heat diffusion and convection in determining an electron transport barrier. They nevertheless provided challenging experimental evidence both for theoretical modelling and for future experiments. (author)

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

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

Heat transport system

International Nuclear Information System (INIS)

Pierce, B.L.

1978-01-01

A heat transport system of small size which can be operated in any orientation consists of a coolant loop containing a vaporizable liquid as working fluid and includes in series a vaporizer, a condenser and two one-way valves and a pressurizer connected to the loop between the two valves. The pressurizer may be divided into two chambers by a flexible diaphragm, an inert gas in one chamber acts as a pneumatic spring for the system. This system is suitable for use in a nuclear-powered artificial heart

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

International Nuclear Information System (INIS)

Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

1977-01-01

The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

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

Integrated heat transport simulation of high ion temperature plasma of LHD

International Nuclear Information System (INIS)

Murakami, S.; Yamaguchi, H.; Sakai, A.

2014-10-01

A first dynamical simulation of high ion temperature plasma with carbon pellet injection of LHD is performed by the integrated simulation GNET-TD + TASK3D. NBI heating deposition of time evolving plasma is evaluated by the 5D drift kinetic equation solver, GNET-TD and the heat transport of multi-ion species plasma (e, H, He, C) is studied by the integrated transport simulation code, TASK3D. Achievement of high ion temperature plasma is attributed to the 1) increase of heating power per ion due to the temporal increase of effective charge, 2) reduction of effective neoclassical transport with impurities, 3) reduction of turbulence transport. The reduction of turbulence transport is most significant contribution to achieve the high ion temperature and the reduction of the turbulent transport from the L-mode plasma (normal hydrogen plasma) is evaluated to be a factor about five by using integrated heat transport simulation code. Applying the Z effective dependent turbulent reduction model we obtain a similar time behavior of ion temperature after the C pellet injection with the experimental results. (author)

Ballistic near-field heat transport in dense many-body systems

Science.gov (United States)

Latella, Ivan; Biehs, Svend-Age; Messina, Riccardo; Rodriguez, Alejandro W.; Ben-Abdallah, Philippe

2018-01-01

Radiative heat transport mediated by near-field interactions is known to be superdiffusive in dilute, many-body systems. Here we use a generalized Landauer theory of radiative heat transfer in many-body planar systems to demonstrate a nonmonotonic transition from superdiffusive to ballistic transport in dense systems. We show that such a transition is associated to a change of the polarization of dominant modes. Our findings are complemented by a quantitative study of the relaxation dynamics of the system in the different regimes of heat transport. This result could have important consequences on thermal management at nanoscale of many-body systems.

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

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.

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

Enhanced heat transport in environmental systems using microencapsulated phase change materials

Science.gov (United States)

Colvin, D. P.; Mulligan, J. C.; Bryant, Y. G.

1992-01-01

A methodology for enhanced heat transport and storage that uses a new two-component fluid mixture consisting of a microencapsulated phase change material (microPCM) for enhanced latent heat transport is outlined. SBIR investigations for NASA, USAF, SDIO, and NSF since 1983 have demonstrated the ability of the two-component microPCM coolants to provide enhancements in heat transport up to 40 times over that of the carrier fluid alone, enhancements of 50 to 100 percent in the heat transfer coefficient, practically isothermal operation when the coolant flow is circulated in an optimal manner, and significant reductions in pump work.

Experimental study on the supercritical startup and heat transport capability of a neon-charged cryogenic loop heat pipe

International Nuclear Information System (INIS)

Guo, Yuandong; Lin, Guiping; He, Jiang; Bai, Lizhan; Zhang, Hongxing; Miao, Jianyin

2017-01-01

Highlights: • A neon-charged CLHP integrated with a G-M cryocooler was designed and investigated. • The CLHP can realize the supercritical startup with an auxiliary heat load of 1.5 W. • Maximum heat transport capability of the CLHP was 4.5 W over a distance of 0.6 m. • There existed an optimum auxiliary heat load to expedite the supercritical startup. • There existed an optimum charged pressure to reach the largest heat transfer limit. - Abstract: Neon-charged cryogenic loop heat pipe (CLHP) can realize efficient cryogenic heat transport in the temperature range of 30–40 K, and promises great application potential in the thermal control of future space infrared exploration system. In this work, extensive experimental studies on the supercritical startup and heat transport capability of a neon-charged CLHP integrated with a G-M cryocooler were carried out, where the effects of the auxiliary heat load applied to the secondary evaporator and charged pressure of the working fluid were investigated. Experimental results showed that the CLHP could successfully realize the supercritical startup with an auxiliary heat load of 1.5 W, and there existed an optimum auxiliary heat load and charged pressure of the working fluid respectively, to achieve the maximum temperature drop rate of the primary evaporator during the supercritical startup. The CLHP could reach a maximum heat transport capability of 4.5 W over a distance of 0.6 m corresponding to the optimum charged pressure of the working fluid; however, the heat transport capability decreased with the increase of the auxiliary heat load. Furthermore, the inherent mechanisms responsible for the phenomena observed in the experiments were analyzed and discussed, to provide a better understanding from the theoretical view.

Network model of free convection within internally heated porous media

International Nuclear Information System (INIS)

Conrad, P.W.

1977-01-01

A hypothetical core-disruptive accident (HCDA) in a liquid metal fast breeder reactor (LMFBR) may result in the formation of an internally heated debris bed. Considerable attention has been given to postulated mechanisms by which such beds may be cooled. It is the purpose of the work described to demonstrate a method for computing the heat transfer from such a bed to the overlying sodium pool due to single-phase, free convection

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

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)

Sodium-water wastage and reactions program performed by general electric in support of the US. AEC LMFBR steam generator development

International Nuclear Information System (INIS)

Greene, D.A.

1975-01-01

This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)

Sodium-water wastage and reactions program performed by general electric in support of the US. AEC LMFBR steam generator development

Energy Technology Data Exchange (ETDEWEB)

Greene, D A

1975-07-01

This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)

Electron heat transport studies using transient phenomena in ASDEX Upgrade

International Nuclear Information System (INIS)

Jacchia, A.; Angioni, C.; Manini, A.; Ryter, F.; Apostoliceanu, M.; Conway, G.; Fahrbach, H.-U.; Kirov, K.K.; Leuterer, F.; Reich, M.; Sutttrop, W.; Cirant, S.; Mantica, P.; De Luca, F.; Weiland, J.

2005-01-01

Experiments in tokamaks suggest that a critical gradient length may cause the resilient behavior of T e profiles, in the absence of ITBs. This agrees in general with ITG/TEM turbulence physics. Experiments in ASDEX Upgrade using modulation techniques with ECH and/or cold pulses demonstrate the existence of a threshold in R/L Te when T e >T i and T e ≤T i . For T e >T i linear stability analyses indicate that electron heat transport is dominated by TEM modes. They agree in the value of the threshold (both T e and n e ) and for the electron heat transport increase above the threshold. The stabilization of TEM modes by collisions yielded by gyro-kinetic calculations, which suggests a transition from TEM to ITG dominated transport at high collisionality, is experimentally demonstrated by comparing heat pulse and steady-state diffusivities. For the T e ∼T i discharges above the threshold the resilience, normalized by T e 3/2 , is similar to that of the TEM dominated cases, despite very different conditions. The heat pinch predicted by fluid modeling of ITG/TEM turbulence is investigated by perturbative transport in off-axis ECH-heated discharges. (author)

Long-distance heat transport by hot water

International Nuclear Information System (INIS)

Munser, H.; Reetz, B.

1990-01-01

From the analysis of the centralized heat supply in the GDR energy-economical and ecological indispensable developments of long-distance heat systems in conurbation are derived. The heat extraction from a nuclear power plant combined with long- distance hot-water transport over about 110 kilometres is investigated and presented as a possibility to perspective base load heat demands for the district around Dresden. By help of industrial-economic, hydraulic and thermic evaluations of first design variants of the transit system the acceptance of this ecologic and energetic preferred solution is proved and requirements for its realization are shown

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

COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

International Nuclear Information System (INIS)

Rhodes, C.A.

1984-12-01

This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

A assessment of loss-of-heat-sink accident with scram in the LMFBR

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Pratt, W.T.; Sun, Y.H.

1978-01-01

A description of a slow core meltdown in a liquid metal fast breeder reactor is presented for conditions of loss-of-heat-sink following neutronic shutdown. Simple models are developed for the prediction of phase changes and/or relocation of the core materials including fuel, clad, ducts, control rod absorber material (B 4 C), and plenum gases. The sequence of events is accounted for and the accident progression is described up to the point of recriticality. The neutronic behavior of the disrupted core is analyzed in R-Z geometry with a static transport theory code. For most scenarios assessed, the reactor is expected to become recritical although large ramp rates are not anticipated. (author)

Assessment of the loss-of-heat-sink accident with scram in the LMFBR

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Pratt, W.T.; Sun, Y.H.

1978-01-01

A description of a slow core meltdown in a liquid metal fast breeder reactor is presented for the conditions of loss-of-heat-sink following neutronic shutdown. Simple models are developed for the prediction of phase changes and/or relocation of the core materials including fuel, clad, ducts, control rod absorber material (B 4 C), and plenum gases. The sequence of events is accounted for and the accident progression is described up to the point of recriticality. The neutronic behavior of the disrupted core is analyzed in R-Z geometry with a static transport theory code. For most scenarios assessed, the reactor is expected to become recritical although large ramp rates are not anticipated

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

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)

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

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

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

Bibliographical survey of heat exchangers for nuclear power plants and problems of HTGR

International Nuclear Information System (INIS)

Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

1977-04-01

The problems in development of heat exchangers for nuclear reactors have been examined in literature survey through Annual Index Subjects of NSA (Nuclear Science Abstracts) for the past ten years. R and D on heat exchangers for LMFBR, HTGR, LWR and HWR are on the increase. In the case of HTGRs, R and D on heat resisting materials including the corrosion and on hydrogen permeation of heat exchanger walls in high temperature pressure helium environment are important. Future R and D subjects for HTGR heat exchangers in showing the high temperature endurance are presented. (auth.)

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

High efficiency heat transport and power conversion system for cascade

International Nuclear Information System (INIS)

Maya, I.; Bourque, R.F.; Creedon, R.L.; Schultz, K.R.

1985-02-01

The Cascade ICF reactor features a flowing blanket of solid BeO and LiAlO 2 granules with very high temperature capability (up to approx. 2300 K). The authors present here the design of a high temperature granule transport and heat exchange system, and two options for high efficiency power conversion. The centrifugal-throw transport system uses the peripheral speed imparted to the granules by the rotating chamber to effect granule transport and requires no additional equipment. The heat exchanger design is a vacuum heat transfer concept utilizing gravity-induced flow of the granules over ceramic heat exchange surfaces. A reference Brayton power cycle is presented which achieves 55% net efficiency with 1300 K peak helium temperature. A modified Field steam cycle (a hybrid Rankine/Brayton cycle) is presented as an alternate which achieves 56% net efficiency

A review on transportation of heat energy over long distance. Exploratory development

Energy Technology Data Exchange (ETDEWEB)

Ma, Q.; Wang, R.Z. [Institute of Refrigeration and Cryogenics, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Luo, L.; Sauce, G. [LOCIE, Polytech' Savoie, Campus Scientifique, Savoie Technolac, 73376 Le Bourget-Du-Lac cedex (France)

2009-08-15

This paper presents a review on transportation of heat energy over long distance. For the transportation of high-temperature heat energy, the chemical catalytic reversible reaction is almost the only way available, and there are several reactions have been studied. For the relatively low-temperature heat energy, which exists widely as waste heat, there are mainly five researching aspects at present: chemical reversible reactions, phase change thermal energy storage and transportation, hydrogen-absorbing alloys, solid-gas adsorption and liquid-gas absorption. The basic principles and the characteristics of these methods are discussed. (author)

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

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

Fluctuation theory for transport properties in multicomponent mixtures: thermodiffusion and heat conductivity

DEFF Research Database (Denmark)

Shapiro, Alexander

2004-01-01

The theory of transport properties in multicomponent gas and liquid mixtures, which was previously developed for diffusion coefficients, is extended onto thermodiffusion coefficients and heat conductivities. The derivation of the expressions for transport properties is based on the general statis...... of the heat conductivity coefficient for ideal gas. (C) 2003 Elsevier B.V. All rights reserved.......The theory of transport properties in multicomponent gas and liquid mixtures, which was previously developed for diffusion coefficients, is extended onto thermodiffusion coefficients and heat conductivities. The derivation of the expressions for transport properties is based on the general...

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

Heat and damp transport in cavity bricks. Waerme- und Feuchtetransport in Hochlochziegeln

Energy Technology Data Exchange (ETDEWEB)

Elsner, M

1987-11-19

The aim of this work is a systematic measurement of the structural effect of cavity bricks on the thermal insulation and thermal storage values depending on the material values of the bricks and the mortar. The arrangement and orientation of the hollow spaces and their dimensions should be varied. Brick shapes with socalled handle slots, which give more convenient handling, and with mortar pockets instead of mortar gaps, should be taken into account in the investigation. Special attention should be paid to the heat transport mechanism in the hollow spaces, where thermal conduction, thermal radiation and convection heat transport are superimposed on one another. The second main aim of the work is the calculation of the coupled heat and damp transport in hollow bricks. The heat and damp transport is described by a coupled system of differential equations, where the decisive transport coefficients should be shown as a function of the variables determining the transport processes. (orig./MM).

Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

International Nuclear Information System (INIS)

Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

1976-01-01

Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

Heat transport the cold way

International Nuclear Information System (INIS)

Anon.

1981-01-01

A novel system for long-distance heat transport is being born in the 'Kernforschungsanlage Juelich' with the project being called 'Nukleare Fernenergie' (nuclear district energy). The project is also known as 'EVA/ADAM' [EVA = Einzelrohr-Versuchs-Anlage (single tube test facility); ADAM = Anlage mit Drei Adiabaten Methanisierungsreaktoren (plant provided with three adiabate methanising reactors)] and is based in principle on transport of energy in chemical bond within a closed loop. In the 60ies already this development was discussed both in the 'Kernforschungsanlage Juelich' and in the 'Rheinische Braunkohlenwerke' independent of each other. In 1975 these two organizations concluded a co-operation contract. (orig.) [de

Investigation of the transportation requirements for fusion power plants

International Nuclear Information System (INIS)

Rhoads, R.E.; Davis, D.K.

1976-09-01

This report presents a general investigation of the transport requirements associated with the construction and operation of conceptual fusion reactors. Projections of amounts of construction and operating materials requiring transportation are presented for several proposed designs. The material to be shipped is described along with the shipping containers that might be used, the transport modes and the expected impact of transporting these materials. Transportation of both radioactive and nonradioactive materials will be required. Most of these materials are routinely shipped by the transportation industry. Transportation requirements of a representative fusion reactor are also compared with Liquid Metal Fast Breeder Reactor (LMFBR) requirements

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

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.

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

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

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

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

Electron heat transport in shaped TCV L-mode plasmas

International Nuclear Information System (INIS)

Camenen, Y; Pochelon, A; Bottino, A; Coda, S; Ryter, F; Sauter, O; Behn, R; Goodman, T P; Henderson, M A; Karpushov, A; Porte, L; Zhuang, G

2005-01-01

Electron heat transport experiments are performed in L-mode discharges at various plasma triangularities, using radially localized electron cyclotron heating to vary independently both the electron temperature T e and the normalized electron temperature gradient R/L T e over a large range. Local gyro-fluid (GLF23) and global collisionless gyro-kinetic (LORB5) linear simulations show that, in the present experiments, trapped electron mode (TEM) is the most unstable mode. Experimentally, the electron heat diffusivity χ e is shown to decrease with increasing collisionality, and no dependence of χ e on R/L T e is observed at high R/L T e values. These two observations are consistent with the predictions of TEM simulations, which supports the fact that TEM plays a crucial role in electron heat transport. In addition, over the broad range of positive and negative triangularities investigated, the electron heat diffusivity is observed to decrease with decreasing plasma triangularity, leading to a strong increase of plasma confinement at negative triangularity

Thaw flow control for liquid heat transport systems

Science.gov (United States)

Kirpich, Aaron S.

1989-01-01

In a liquid metal heat transport system including a source of thaw heat for use in a space reactor power system, the thaw flow throttle or control comprises a fluid passage having forward and reverse flow sections and a partition having a plurality of bleed holes therein to enable fluid flow between the forward and reverse sections. The flow throttle is positioned in the system relatively far from the source of thaw heat.

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

Perturbative Heat Transport Experiments on TJ-II

International Nuclear Information System (INIS)

Eguilor, S.; Castejon, F.; Luna, E. de la; Cappa, A.; Likin, K.; Fernandez, A.; Tj-II, T.

2002-01-01

Heat wave experiments are performed on TJ-II stellarator plasmas to estimate both heat diffusivity and power deposition profiles. High frequency ECRH modulation experiments are used to obtain the power deposition profiles, which is observed to be wider and duller than estimated by tracing techniques. The causes of this difference are discussed in the paper. Fourier analysis techniques are used to estimate the heat diffusivity in low frequency ECRH modulation experiments. This include the power deposition profile as a new ingredient. ECHR switch on/off experiments are exploited to obtain power deposition and heat diffusivities profile. Those quantities are compared with the obtained by modulation experiments and transport analysis, showing a good agreement. (Author) 18 refs

Perturbative Heat Transport Experiments on TJ-II

Energy Technology Data Exchange (ETDEWEB)

Eguilor, S.; Castejon, F.; Luna, E. de la; Cappa, A.; Likin, K.; Fernandez, A.; Tj-II, T.

2002-07-01

Heat wave experiments are performed on TJ-II stellarator plasmas to estimate both heat diffusivity and power deposition profiles. High frequency ECRH modulation experiments are used to obtain the power deposition profiles, which is observed to be wider and duller than estimated by tracing techniques. The causes of this difference are discussed in the paper. Fourier analysis techniques are used to estimate the heat diffusivity in low frequency ECRH modulation experiments. This include the power deposition profile as a new ingredient. ECHR switch on/off experiments are exploited to obtain power deposition and heat diffusivities profile. Those quantities are compared with the obtained by modulation experiments and transport analysis, showing a good agreement. (Author) 18 refs.

Diffusive and convective transport modelling from analysis of ECRH-stimulated electron heat wave propagation. [ECRH (Electron Cyclotron Resonance Heating)

Energy Technology Data Exchange (ETDEWEB)

Erckmann, V; Gasparino, U; Giannone, L. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)) (and others)

1992-01-01

ECRH power modulation experiments in toroidal devices offer the chance to analyze the electron heat transport more conclusively: the electron heat wave propagation can be observed by ECE (or SX) leading to radial profiles of electron temperature modulation amplitude and time delay (phase shift). Taking also the stationary power balance into account, the local electron heat transport can be modelled by a combination of diffusive and convective transport terms. This method is applied to ECRH discharges in the W7-AS stellarator (B=2.5T, R=2m, a[<=]18 cm) where the ECRH power deposition is highly localized. In W7-AS, the T[sub e] modulation profiles measured by a high resolution ECE system are the basis for the local transport analysis. As experimental errors limit the separation of diffusive and convective terms in the electron heat transport for central power deposition, also ECRH power modulation experiments with off-axis deposition and inward heat wave propagation were performed (with 70 GHz o-mode as well as with 140 GHz x-mode for increased absorption). Because collisional electron-ion coupling and radiative losses are only small, low density ECRH discharges are best candidates for estimating the electron heat flux from power balance. (author) 2 refs., 3 figs.

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

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

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)

Non-standard model for electron heat transport for multidimensional hydrodynamic codes

Energy Technology Data Exchange (ETDEWEB)

Nicolai, Ph.; Busquet, M.; Schurtz, G. [CEA/DAM-Ile de France, 91 - Bruyeres Le Chatel (France)

2000-07-01

In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

Non-standard model for electron heat transport for multidimensional hydrodynamic codes

International Nuclear Information System (INIS)

Nicolai, Ph.; Busquet, M.; Schurtz, G.

2000-01-01

In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

Magnetically Modulated Heat Transport in a Global Simulation of Solar Magneto-convection

Energy Technology Data Exchange (ETDEWEB)

Cossette, Jean-Francois [Laboratory for Atmospheric and Space Physics, Campus Box 600, University of Colorado, Boulder, CO 80303 (United States); Charbonneau, Paul [Département de Physique, Université de Montréal, C.P. 6128, Succ. Centre-Ville, Montréal, QC H3C 3J7 (Canada); Smolarkiewicz, Piotr K. [European Centre for Medium-Range Weather Forecasts, Reading, RG2 9AX (United Kingdom); Rast, Mark P., E-mail: Jean-Francois.Cossette@lasp.colorado.edu, E-mail: paulchar@astro.umontreal.ca, E-mail: smolar@ecmwf.int, E-mail: Mark.Rast@lasp.colorado.edu [Department of Astrophysical and Planetary Sciences, Laboratory for Atmospheric and Space Physics, Campus Box 391, University of Colorado, Boulder, CO 80303 (United States)

2017-05-20

We present results from a global MHD simulation of solar convection in which the heat transported by convective flows varies in-phase with the total magnetic energy. The purely random initial magnetic field specified in this experiment develops into a well-organized large-scale antisymmetric component undergoing hemispherically synchronized polarity reversals on a 40 year period. A key feature of the simulation is the use of a Newtonian cooling term in the entropy equation to maintain a convectively unstable stratification and drive convection, as opposed to the specification of heating and cooling terms at the bottom and top boundaries. When taken together, the solar-like magnetic cycle and the convective heat flux signature suggest that a cyclic modulation of the large-scale heat-carrying convective flows could be operating inside the real Sun. We carry out an analysis of the entropy and momentum equations to uncover the physical mechanism responsible for the enhanced heat transport. The analysis suggests that the modulation is caused by a magnetic tension imbalance inside upflows and downflows, which perturbs their respective contributions to heat transport in such a way as to enhance the total convective heat flux at cycle maximum. Potential consequences of the heat transport modulation for solar irradiance variability are briefly discussed.

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

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

Validation of an intermediate heat exchanger model for real time analysis

International Nuclear Information System (INIS)

Tzanos, C.P.

1986-11-01

A new method was presented for LMFBR intermediate heat exchanger (IHX) analysis in real time for purposes of continuous on-line data validation, plant state verification and fault identification. For the validation of this methodology the EBR-II IHX transient during Test 8A was analyzed. This paper presents the results of this analysis

Heat transport analysis in a district heating and snow melting system in Sapporo and Ishikari, Hokkaido applying waste heat from GTHTR300

International Nuclear Information System (INIS)

Kasahara, Seiji; Kamiji, Yu; Terada, Atsuhiko; Yan Xing; Inagaki, Yoshiyuki; Murata, Tetsuya; Mori, Michitsugu

2015-01-01

A district heating and snow melting system utilizing waste heat from Gas Turbine High temperature Gas Reactor of 300 MW_e (GTHTR300), a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. Application areas are set in Sapporo and Ishikari, the heavy snowfall cities in Northern Japan. The heat transport analyses are carried out by modeling the components in the system; pipelines of the secondary water loops between GTHTR300s and heat demand district and heat exchangers to transport the heat from the secondary water loops to the tertiary loops in the district. Double pipe for the secondary loops are advantageous for less heat loss and smaller excavation area. On the other hand, these pipes has disadvantage of more electricity consumption for pumping. Most of the heat demand in the month of maximum requirement can be supplied by 2 GTHTR300s and delivered by 9 secondary loops and around 5000 heat exchangers. Closer location of GTHTR300 site to the heat demand district is largely advantageous economically. Less decrease of the distance from 40 km to 20 km made the heat loss half and cost of the heat transfer system 22% smaller. (author)

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

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

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

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)

Thermal evaluation facility for LMFBR spent fuel transport

International Nuclear Information System (INIS)

Wesley, D.A.

1980-04-01

A full-scale mock-up of a 217 pin breeder reactor fuel assembly in a cylindrical pipe was initially designed and constructed by Oak Ridge National Laboratory (ORNL). It was transferred to Sandia where it was extensively redesigned and modified. The 217 pin hexagonal core assembly was installed in a smaller diameter stainless steel pipe which more closely represents the diameter of a shipping canister or shipping cask basket wall. Two-hundred four of the tubes are electrically heated over an active length of 4-feet and the remaining thirteen are instrumented with multiple junction thermocouples which can be traversed axially. Thermocouples and heat-flux gauges are located on the hex core and canister perimeters at several axial locations

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

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

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.)

Heat Transport in Gapped Spin-Chain Systems

International Nuclear Information System (INIS)

Shimshoni, E.

2006-01-01

Full Text: We study the contribution of magnetic excitations to the heat transport in gapped spin-chain systems. These systems are characterized by a substantially enhanced heat conductivity, which can be traced back to the existence of weakly violated conservation laws. We focus particularly on the behavior of clean two-leg spin ladder compounds, where one-dimensional exotic spin excitations are coupled to three-dimensional phonons. We show that the contributions of the two types of heat carriers can not be easily disentangled. Depending on the ratios of spin gaps and the Debye energy, the heat conductivity can be either exponentially increasing or exponentially decreasing as a function of temperature (T). In addition, the magnetic contribution to the total heat conductivity may be either positive or negative. We discuss its T-dependence in various possible regimes, and note that in most regimes it is dominated by spin-phonon drag: the two types of heat carriers have almost the

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)

Subcooled He II heat transport in the channel with abrupt contractions/enlargements

International Nuclear Information System (INIS)

Maekawa, R.; Iwamoto, A.; Hamaguchi, S.; Mito, T.

2002-01-01

Heat transport mechanisms for subcooled He II in the channel with abrupt contractions and/or enlargements have been investigated under steady state conditions. The channel, made of G-10, contains various contraction geometries to simulate the cooling channel of a superconducting magnet. In other words, contractions are periodically placed along the channel to simulate the spacers within the magnet winding. A copper block heater inputs the heat to the channel from one end, while the other end is open to the He II bath. Temperature profiles were measured with temperature sensors embedded in the channel as a function of heat input. Calculations were performed using a simple one-dimensional turbulent heat transport equation and with geometric factor consideration. The effects on heat transport mechanisms in He II caused by abrupt change of channel geometry and size are discussed

Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Ribando, R.J.

1979-01-01

A comparison is made between computed results and experimental data for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL) used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility using a variety of initial conditions and testing parameters. Specifically, in this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow. The computer program used to analyze the test, LONAC (LOw flow and NAtural Convection) is an ORNL-developed, fast-running, one-dimensional, single-phase, finite-difference model used for simulating forced and free convection transients in the THORS loop.

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

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

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

Selection of engineering materials and fabrication of liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Patriarca, P.

1975-01-01

Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

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)

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.)

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)

Comparison of temperature estimates from heat transport model and electrical resistivity tomography during a shallow heat injection and storage experiment

OpenAIRE

Hermans, Thomas; Daoudi, Moubarak; Vandenbohede, Alexander; Robert, Tanguy; Caterina, David; Nguyen, Frédéric

2012-01-01

Groundwater resources are increasingly used around the world as geothermal systems. Understanding physical processes and quantification of parameters determining heat transport in porous media is therefore important. Geophysical methods may be useful in order to yield additional information with greater coverage than conventional wells. We report a heat transport study during a shallow heat injection and storage field test. Heated water (about 50°C) was injected for 6 days at the rate of 80 l...

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

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

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)

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

Molecular dynamics study on heat transport from single-walled carbon nanotubes to Si substrate

Energy Technology Data Exchange (ETDEWEB)

Feng, Ya; Zhu, Jie, E-mail: zhujie@iet.cn; Tang, Da-Wei

2015-02-06

In this paper, non-equilibrium molecular dynamics simulations were performed to investigate the heat transport between a vertically aligned single-walled carbon nanotube (SWNT) and Si substrate, to find out the influence of temperature and system sizes, including diameter and length of SWNT and measurements of substrate. Results revealed that high temperature hindered heat transport in SWNT itself but was a beneficial stimulus for heat transport at interface of SWNT and Si. Furthermore, the system sizes strongly affected the peaks in vibrational density of states of Si, which led to interfacial thermal conductance dependent on system sizes. - Highlights: • NEMD is performed to simulate the heat transport from SWNT to Si substrate. • We analyze both interfacial thermal conductance and thermal conductivity of SWNT. • High temperature is a beneficial stimulus for heat transport at the interface. • Interfacial thermal conductance strongly depends on the sizes of SWNT and substrate. • We calculate VDOS of C and Si atoms to analyze phonon couplings between them.

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.)

Circum-Antarctic Shoreward Heat Transport Derived From an Eddy- and Tide-Resolving Simulation

Science.gov (United States)

Stewart, Andrew L.; Klocker, Andreas; Menemenlis, Dimitris

2018-01-01

Almost all heat reaching the bases of Antarctica's ice shelves originates from warm Circumpolar Deep Water in the open Southern Ocean. This study quantifies the roles of mean and transient flows in transporting heat across almost the entire Antarctic continental slope and shelf using an ocean/sea ice model run at eddy- and tide-resolving (1/48°) horizontal resolution. Heat transfer by transient flows is approximately attributed to eddies and tides via a decomposition into time scales shorter than and longer than 1 day, respectively. It is shown that eddies transfer heat across the continental slope (ocean depths greater than 1,500 m), but tides produce a stronger shoreward heat flux across the shelf break (ocean depths between 500 m and 1,000 m). However, the tidal heat fluxes are approximately compensated by mean flows, leaving the eddy heat flux to balance the net shoreward heat transport. The eddy-driven cross-slope overturning circulation is too weak to account for the eddy heat flux. This suggests that isopycnal eddy stirring is the principal mechanism of shoreward heat transport around Antarctica, though likely modulated by tides and surface forcing.

Characteristics of convective heat transport in a packed pebble-bed reactor

Energy Technology Data Exchange (ETDEWEB)

Abdulmohsin, Rahman S., E-mail: rsar62@mst.edu [Department of Chemical and Biochemical Engineering, Missouri University of Science and Technology, 400 West 11th Street/231 Schrenk Hall, Rolla, MO 65409-1230 (United States); Al-Dahhan, Muthanna H., E-mail: aldahhanm@mst.edu [Department of Chemical and Biochemical Engineering, Missouri University of Science and Technology, 400 West 11th Street/231 Schrenk Hall, Rolla, MO 65409-1230 (United States); Department of Nuclear Engineering, 301 W. 14th St./222 Fulton Hall (United States)

2015-04-01

Highlights: • A fast-response heat transfer probe has been developed and used in this work. • Heat transport has been quantified in terms of local heat transfer coefficients. • The method of the electrically heated single sphere in packing has been applied. • The heat transfer coefficient increases from the center to the wall of packed bed. • This work advancing the knowledge of heat transport in the studied packed bed. - Abstract: Obtaining more precise results and a better understanding of the heat transport mechanism in the dynamic core of packed pebble-bed reactors is needed because this mechanism poses extreme challenges to the reliable design and efficient operation of these reactors. This mechanism can be quantified in terms of a solid-to-gas convective heat transfer coefficient. Therefore, in this work, the local convective heat transfer coefficients and their radial profiles were measured experimentally in a separate effect pilot-plant scale and cold-flow experimental setup of 0.3 m in diameter, using a sophisticated noninvasive heat transfer probe of spherical type. The effect of gas velocity on the heat transfer coefficient was investigated over a wide range of Reynolds numbers of practical importance. The experimental investigations of this work include various radial locations along the height of the bed. It was found that an increase in coolant gas flow velocity causes an increase in the heat transfer coefficient and that effect of the gas flow rate varies from laminar to turbulent flow regimes at all radial positions of the studied packed pebble-bed reactor. The results show that the local heat transfer coefficient increases from the bed center to the wall due to the change in the bed structure, and hence, in the flow pattern of the coolant gas. The findings clearly indicate that one value of an overall heat transfer coefficient cannot represent the local heat transfer coefficients within the bed; therefore, correlations are needed to

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)

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.)

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

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

A low-frequency wave motion mechanism enables efficient energy transport in carbon nanotubes at high heat fluxes.

Science.gov (United States)

Zhang, Xiaoliang; Hu, Ming; Poulikakos, Dimos

2012-07-11

The great majority of investigations of thermal transport in carbon nanotubes (CNTs) in the open literature focus on low heat fluxes, that is, in the regime of validity of the Fourier heat conduction law. In this paper, by performing nonequilibrium molecular dynamics simulations we investigated thermal transport in a single-walled CNT bridging two Si slabs under constant high heat flux. An anomalous wave-like kinetic energy profile was observed, and a previously unexplored, wave-dominated energy transport mechanism is identified for high heat fluxes in CNTs, originated from excited low frequency transverse acoustic waves. The transported energy, in terms of a one-dimensional low frequency mechanical wave, is quantified as a function of the total heat flux applied and is compared to the energy transported by traditional Fourier heat conduction. The results show that the low frequency wave actually overtakes traditional Fourier heat conduction and efficiently transports the energy at high heat flux. Our findings reveal an important new mechanism for high heat flux energy transport in low-dimensional nanostructures, such as one-dimensional (1-D) nanotubes and nanowires, which could be very relevant to high heat flux dissipation such as in micro/nanoelectronics applications.

Mobile heat accumulators for lorry or train transport?; Mobile Waermespeicher fuer den LKW- oder Zugtransport?

Energy Technology Data Exchange (ETDEWEB)

Goldenberg, Philipp

2013-07-01

Where heat grids cannot be laid for geographic reasons, mobile heat accumulators may be appropriate. The mobile heat accumulators are transported by lorry or train between the heat source and the heat sink. The waste heat can be decoupled from biogas plants, waste incineration plants or industrial sites. Existing road or rail networks can be used for transportation. Decisive factors to achieve low heat production costs are: free waste heat, large and continuous heat quantities as well as a short distance between the heat source and the heat sink. (orig.)

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

Stable solutions of nonlocal electron heat transport equations

International Nuclear Information System (INIS)

Prasad, M.K.; Kershaw, D.S.

1991-01-01

Electron heat transport equations with a nonlocal heat flux are in general ill-posed and intrinsically unstable, as proved by the present authors [Phys. Fluids B 1, 2430 (1989)]. A straightforward numerical solution of these equations will therefore lead to absurd results. It is shown here that by imposing a minimal set of constraints on the problem it is possible to arrive at a globally stable, consistent, and energy conserving numerical solution

The adjoint space in heat transport theory

International Nuclear Information System (INIS)

Dam, H. van; Hoogenboom, J.E.

1980-01-01

The mathematical concept of adjoint operators is applied to the heat transport equation and an adjoint equation is defined with a detector function as source term. The physical meaning of the solutions for the latter equation is outlined together with an application in the field of perturbation analysis. (author)

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

Latent heat increases storage capacity. Heat transport by truck; Latente warmte vergroot opslagcapaciteit. Warmtetransport per vrachtauto is soms heel slim

Energy Technology Data Exchange (ETDEWEB)

De Jong, K.

2012-11-15

The project-group Biomass CHP (combined production of heat and power) organized a tour with a workshop in Dortmund, Germany, September 26, 2012, on storage and transport of heat and biogas. There are several projects in Germany involving road transport of heat by means of containers. A swimming pool in Dortmund already is using this option since 2008. Waste heat from a CHP-installation for landfill gas is collected from a waste dump [Dutch] De projectgroep Biomassa en WKK organiseerde 26 September een excursie met workshop in Dortmund over opslag en transport van warmte en biogas. Er zijn in Duitsland al meerdere projecten waarbij warmte per container over de weg wordt vervoerd. Een Dortmunds zwembad werkt hier al sinds 2008 mee. De restwarmte van een wkk op stortgas wordt opgehaald bij een afvalstortplaats.

Heat transport in low-dimensional materials: A review and perspective

Directory of Open Access Journals (Sweden)

Zhiping Xu

2016-05-01

Full Text Available Heat transport is a key energetic process in materials and devices. The reduced sample size, low dimension of the problem and the rich spectrum of material imperfections introduce fruitful phenomena at nanoscale. In this review, we summarize recent progresses in the understanding of heat transport process in low-dimensional materials, with focus on the roles of defects, disorder, interfaces, and the quantum-mechanical effect. New physics uncovered from computational simulations, experimental studies, and predictable models will be reviewed, followed by a perspective on open challenges.

Integral representation of nonlinear heat transport

International Nuclear Information System (INIS)

Kishimoto, Y.; Mima, K.; Haines, M.G.

1985-07-01

The electron distribution function in a plasma with steep temperature gradient is obtained from a Fokker-Planck equation by Green's function method. The formula describes the nonlocal effects on thermal transport over the range, λ e /L e /L → 0. As an example, the heat wave is analyzed numerically by the integral formula and it is found that the previous simulation results are well reproduced. (author)

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

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

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

Moment approach to neoclassical flows, currents and transport in auxiliary heated tokamaks

International Nuclear Information System (INIS)

Kim, Yil Bong.

1988-02-01

The moment approach is utilized to derive the full complement of neoclassical transport processes in auxiliary heated tokamaks. The effects of auxiliary heating [neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH)] considered arise from the collisional interaction between the background plasma species and the fast-ion-tail species. From a known fast ion distribution function we evaluate the parallel (to the magnetic field) momentum and heat flow inputs to the background plasma. Then, through the momentum and heat flow balance equations, we can determine the induced parallel flows (and current) and radial transpot fluxes in ''equilibrium'' (on the time scale much longer than the collisional relaxation time, i.e., t >> 1ν/sub ii/). In addition to the fast-ion-induced current, the total neoclassical current includes the boostap current, which is driven by the pressure and temperature gradients, the Pfirsch-Schlueter current which is required for charge neutrality, and the neoclassical (including trapped particle effects) Spitzer current due to the parallel electric field. The radial transport fluxes also include off-diagonal compnents in the transport matrix which correspond to the Ware (neoclassical) pinch due to the inductive applied electric field an the fast-ion-induced radial fluxes, in addition to the usual pressure- and temperature-gradient-driven fluxes (particle diffusion and heat conduction). Once the tranport coefficient are completely determined, the radial fluxes and the heat fluxes can be substituted into the density and energy evolution equations to provide a complete description of ''equilibrium'' (δδt << ν/sub ii/) neoclassical transport processes in a plasma. 47 refs., 14 figs

The Role of Ocean and Atmospheric Heat Transport in the Arctic Amplification

Science.gov (United States)

Vargas Martes, R. M.; Kwon, Y. O.; Furey, H. H.

2017-12-01

Observational data and climate model projections have suggested that the Arctic region is warming around twice faster than the rest of the globe, which has been referred as the Arctic Amplification (AA). While the local feedbacks, e.g. sea ice-albedo feedback, are often suggested as the primary driver of AA by previous studies, the role of meridional heat transport by ocean and atmosphere is less clear. This study uses the Community Earth System Model version 1 Large Ensemble simulation (CESM1-LE) to seek deeper understanding of the role meridional oceanic and atmospheric heat transports play in AA. The simulation consists of 40 ensemble members with the same physics and external forcing using a single fully coupled climate model. Each ensemble member spans two time periods; the historical period from 1920 to 2005 using the Coupled Model Intercomparison Project Phase 5 (CMIP5) historical forcing and the future period from 2006 to 2100 using the CMIP5 Representative Concentration Pathways 8.5 (RCP8.5) scenario. Each of the ensemble members are initialized with slightly different air temperatures. As the CESM1-LE uses a single model unlike the CMIP5 multi-model ensemble, the internal variability and the externally forced components can be separated more clearly. The projections are calculated by comparing the period 2081-2100 relative to the time period 2001-2020. The CESM1-LE projects an AA of 2.5-2.8 times faster than the global average, which is within the range of those from the CMIP5 multi-model ensemble. However, the spread of AA from the CESM1-LE, which is attributed to the internal variability, is 2-3 times smaller than that of the CMIP5 ensemble, which may also include the inter-model differences. CESM1LE projects a decrease in the atmospheric heat transport into the Arctic and an increase in the oceanic heat transport. The atmospheric heat transport is further decomposed into moisture transport and dry static energy transport. Also, the oceanic heat

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.)

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

Climate in the Absence of Ocean Heat Transport

Science.gov (United States)

Rose, B. E. J.

2015-12-01

The energy transported by the oceans to mid- and high latitudes is small compared to the atmosphere, yet exerts an outsized influence on the climate. A key reason is the strong interaction between ocean heat transport (OHT) and sea ice extent. I quantify this by comparing a realistic control climate simulation with a slab ocean simulation in which OHT is disabled. Using the state-of-the-art CESM with a realistic present-day continental configuration, I show that the absence of OHT leads to a 23 K global cooling and massive expansion of sea ice to near 30º latitude in both hemisphere. The ice expansion is asymmetric, with greatest extent in the South Pacific and South Indian ocean basins. I discuss implications of this enormous and asymmetric climate change for atmospheric circulation, heat transport, and tropical precipitation. Parameter sensitivity studies show that the simulated climate is far more sensitive to small changes in ice surface albedo in the absence of OHT, with some perturbations sufficient to cause a runaway Snowball Earth glaciation. I conclude that the oceans are responsible for an enormous global warming by mitigating an otherwise very potent sea ice albedo feedback, but that the magnitude of this effect is still rather uncertain. I will also present some ideas on adapting the simple energy balance model to account for the enhanced sensitivity of sea ice to heating from the ocean.

Role of ocean heat transport in climates of tidally locked exoplanets around M dwarf stars.

Science.gov (United States)

Hu, Yongyun; Yang, Jun

2014-01-14

The distinctive feature of tidally locked exoplanets is the very uneven heating by stellar radiation between the dayside and nightside. Previous work has focused on the role of atmospheric heat transport in preventing atmospheric collapse on the nightside for terrestrial exoplanets in the habitable zone around M dwarfs. In the present paper, we carry out simulations with a fully coupled atmosphere-ocean general circulation model to investigate the role of ocean heat transport in climate states of tidally locked habitable exoplanets around M dwarfs. Our simulation results demonstrate that ocean heat transport substantially extends the area of open water along the equator, showing a lobster-like spatial pattern of open water, instead of an "eyeball." For sufficiently high-level greenhouse gases or strong stellar radiation, ocean heat transport can even lead to complete deglaciation of the nightside. Our simulations also suggest that ocean heat transport likely narrows the width of M dwarfs' habitable zone. This study provides a demonstration of the importance of exooceanography in determining climate states and habitability of exoplanets.

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.

Diffusive and convective transport modelling from analysis of ECRH-stimulated electron heat wave propagation

International Nuclear Information System (INIS)

Erckmann, V.; Gasparino, U.; Giannone, L.

1992-01-01

ECRH power modulation experiments in toroidal devices offer the chance to analyze the electron heat transport more conclusively: the electron heat wave propagation can be observed by ECE (or SX) leading to radial profiles of electron temperature modulation amplitude and time delay (phase shift). Taking also the stationary power balance into account, the local electron heat transport can be modelled by a combination of diffusive and convective transport terms. This method is applied to ECRH discharges in the W7-AS stellarator (B=2.5T, R=2m, a≤18 cm) where the ECRH power deposition is highly localized. In W7-AS, the T e modulation profiles measured by a high resolution ECE system are the basis for the local transport analysis. As experimental errors limit the separation of diffusive and convective terms in the electron heat transport for central power deposition, also ECRH power modulation experiments with off-axis deposition and inward heat wave propagation were performed (with 70 GHz o-mode as well as with 140 GHz x-mode for increased absorption). Because collisional electron-ion coupling and radiative losses are only small, low density ECRH discharges are best candidates for estimating the electron heat flux from power balance. (author) 2 refs., 3 figs

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)

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

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

Turbulent transport regimes and the SOL heat flux width

Science.gov (United States)

Myra, J. R.; D'Ippolito, D. A.; Russell, D. A.

2014-10-01

Understanding the responsible mechanisms and resulting scaling of the scrape-off layer (SOL) heat flux width is important for predicting viable operating regimes in future tokamaks, and for seeking possible mitigation schemes. Simulation and theory results using reduced edge/SOL turbulence models have produced SOL widths and scalings in reasonable accord with experiments in many cases. In this work, we attempt to qualitatively and conceptually understand various regimes of edge/SOL turbulence and the role of turbulent transport in establishing the SOL heat flux width. Relevant considerations include the type and spectral characteristics of underlying instabilities, the location of the gradient drive relative to the SOL, the nonlinear saturation mechanism, and the parallel heat transport regime. Recent SOLT turbulence code results are employed to understand the roles of these considerations and to develop analytical scalings. We find a heat flux width scaling with major radius R that is generally positive, consistent with older results reviewed in. The possible relationship of turbulence mechanisms to the heuristic drift mechanism is considered, together with implications for future experiments. Work supported by US DOE grant DE-FG02-97ER54392.

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.)

Magnetic-field asymmetry of nonlinear thermoelectric and heat transport

International Nuclear Information System (INIS)

Hwang, Sun-Yong; Sánchez, David; López, Rosa; Lee, Minchul

2013-01-01

Nonlinear transport coefficients do not obey, in general, reciprocity relations. We here discuss the magnetic-field asymmetries that arise in thermoelectric and heat transport of mesoscopic systems. Based on a scattering theory of weakly nonlinear transport, we analyze the leading-order symmetry parameters in terms of the screening potential response to either voltage or temperature shifts. We apply our general results to a quantum Hall antidot system. Interestingly, we find that certain symmetry parameters show a dependence on the measurement configuration. (paper)

Latent heat transport and microlayer evaporation in nucleate boiling

International Nuclear Information System (INIS)

Jawurek, H.H.

1977-08-01

Part 1 of this work provides a broad overview and, where possible, a quantitative assessment of the complex physical processes which together constitute the mechanism of nucleate boiling heat transfer. It is shown that under a wide range of conditions the primary surface-to-liquid heat flows within an area of bubble influence are so redistributed as to manifest themselves predominantly as latent heat transport, that is, as vaporisation into attached bubbles. Part 2 deals in greater detail with one of the component processes of latent heat transport, namely microlayer evaporation. A literature review reveals the need for synchronised records of microlayer geometry versus time and of normal bubble growth and departure. An apparatus developed to provide such records is described. High-speed cine interference photography from beneath and through a transparent heating surface provided details of microlayer geometry and an image reflection system synchronised these records with the bubble profile views. Results are given for methanol and ethanol boiling at sub-atmospheric pressures and at various heat fluxes and bulk subcoolings. In all cases it is found that microlayers were of sub-micron thickness, that microlayer thinning was restricted to the inner layer edge (with the thickness elsewhere remaining constant or increasing with time) and that the contribution of this visible evaporation to the total vapour flow into bubbles was negligible. The observation of thickening towards the outer microlayer edge, however, demonstrates that a liquid replenishment flow occurred simultaneously with the evaporation process

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.

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

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)

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

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

Study of electronic heat transport in plasma through diagnosis based on modulated electron cyclotron heating; Etudes de transport de la chaleur electronique par injection modulee d'ondes a la frequence cyclotronique electronique

Energy Technology Data Exchange (ETDEWEB)

Clemencon, A.; Guivarch, C

2003-07-01

In order to make nuclear fusion energetically profitable, it is crucial to heat and confine the plasma efficiently. Studying the behavior of the heat diffusion coefficient is a key issue in this matter. The use of modulated electron cyclotron heating as a diagnostic has suggested the existence of a transport barrier under certain plasma conditions. We have determined the solution to the heat transport equation, for several heat diffusion coefficient profiles. By comparing the analytical solutions with experimental data; we are able to study the heat diffusion coefficient profile. Thus, in certain experiments, we can confirm that the heat diffusion coefficient switches from low to high values at the radius where the electron cyclotron heat deposition is made. (authors)

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

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

Thermophysical and heat transfer properties of phase change material candidate for waste heat transportation system

Science.gov (United States)

Kaizawa, Akihide; Maruoka, Nobuhiro; Kawai, Atsushi; Kamano, Hiroomi; Jozuka, Tetsuji; Senda, Takeshi; Akiyama, Tomohiro

2008-05-01

A waste heat transportation system trans-heat (TH) system is quite attractive that uses the latent heat of a phase change material (PCM). The purpose of this paper is to study the thermophysical properties of various sugars and sodium acetate trihydrate (SAT) as PCMs for a practical TH system and the heat transfer property between PCM selected and heat transfer oil, by using differential scanning calorimetry (DSC), thermogravimetry-differential thermal analysis (TG-DTA) and a heat storage tube. As a result, erythritol, with a large latent heat of 344 kJ/kg at melting point of 117°C, high decomposition point of 160°C and excellent chemical stability under repeated phase change cycles was found to be the best PCM among them for the practical TH system. In the heat release experiments between liquid erythritol and flowing cold oil, we observed foaming phenomena of encapsulated oil, in which oil droplet was coated by solidification of PCM.

First-principles simulations of heat transport

Science.gov (United States)

Puligheddu, Marcello; Gygi, Francois; Galli, Giulia

2017-11-01

Advances in understanding heat transport in solids were recently reported by both experiment and theory. However an efficient and predictive quantum simulation framework to investigate thermal properties of solids, with the same complexity as classical simulations, has not yet been developed. Here we present a method to compute the thermal conductivity of solids by performing ab initio molecular dynamics at close to equilibrium conditions, which only requires calculations of first-principles trajectories and atomic forces, thus avoiding direct computation of heat currents and energy densities. In addition the method requires much shorter sequential simulation times than ordinary molecular dynamics techniques, making it applicable within density functional theory. We discuss results for a representative oxide, MgO, at different temperatures and for ordered and nanostructured morphologies, showing the performance of the method in different conditions.

Heat transport in an anharmonic crystal

Science.gov (United States)

Acharya, Shiladitya; Mukherjee, Krishnendu

2018-04-01

We study transport of heat in an ordered, anharmonic crystal in the form of slab geometry in three dimensions. Apart from attaching baths of Langevin type to two extreme surfaces, we also attach baths of same type to the intermediate surfaces of the slab. Since the crystal is uninsulated, it exchanges energy with the intermediate heat baths. We find that both Fourier’s law of heat conduction and the Newton’s law of cooling hold to leading order in anharmonic coupling. The leading behavior of the temperature profile is exponentially falling from high to low temperature surface of the slab. As the anharmonicity increases, profiles fall more below the harmonic one in the log plot. In the thermodynamic limit thermal conductivity remains independent of the environment temperature and its leading order anharmonic contribution is linearly proportional to the temperature change between the two extreme surfaces of the slab. A fast crossover from one-dimensional (1D) to three-dimensional (3D) behavior of the thermal conductivity is observed in the system.

Mesoscale Eddies in the Northwestern Pacific Ocean: Three-Dimensional Eddy Structures and Heat/Salt Transports

Science.gov (United States)

Dong, Di; Brandt, Peter; Chang, Ping; Schütte, Florian; Yang, Xiaofeng; Yan, Jinhui; Zeng, Jisheng

2017-12-01

The region encompassing the Kuroshio Extension (KE) in the Northwestern Pacific Ocean (25°N-45°N and 130°E-180°E) is one of the most eddy-energetic regions of the global ocean. The three-dimensional structures and transports of mesoscale eddies in this region are comprehensively investigated by combined use of satellite data and Argo profiles. With the allocation of Argo profiles inside detected eddies, the spatial variations of structures of eddy temperature and salinity anomalies are analyzed. The results show that eddies predominantly have subsurface (near-surface) intensified temperature and salinity anomalies south (north) of the KE jet, which is related to different background stratifications between these regions. A new method based on eddy trajectories and the inferred three-dimensional eddy structures is proposed to estimate heat and salt transports by eddy movements in a Lagrangian framework. Spatial distributions of eddy transports are presented over the vicinity of the KE for the first time. The magnitude of eddy-induced meridional heat (freshwater volume) transport is on the order of 0.01 PW (103 m3/s). The eddy heat transport divergence results in an oceanic heat loss south and heat gain north of the KE, thereby reinforcing and counteracting the oceanic heat loss from air-sea fluxes south and north of the KE jet, respectively. It also suggests a poleward heat transport across the KE jet due to eddy propagation.

Solar-energy heats a transportation test center--Pueblo, Colorado

Science.gov (United States)

1981-01-01

Petroleum-base, thermal energy transport fluid circulating through 583 square feet of flat-plate solar collectors accumulates majority of energy for space heating and domestic hot-water of large Test Center. Report describes operation, maintenance, and performance of system which is suitable for warehouses and similar buildings. For test period from February 1979 to January 1980, solar-heating fraction was 31 percent, solar hot-water fraction 79 percent.

Experimental study on fast neutron streaming through grid-plate shield of a LMFBR

International Nuclear Information System (INIS)

Oka, Yoshiaki; Wakabayashi, Hiroaki; An, Shigehiro; Suzuki, Ikunori.

1976-01-01

Neutron streaming through the holes penetrating the grid plate shield of a prototype LMFBR was experimentally examined. The mockups of the grid plate shield were made of iron and aluminum. Experiments were conducted at the vertical column of ''YAYOI'', the fast neutron source reactor of University of Tokyo. A He-3 spectrometer was employed in order to measure the transmitted neutron spectrum, while rhodium and indium threshold foils were for the integral flux above specific energies and their spatial distributions in the form of reaction rates. The streaming factor for usual small bended holes is 1.28+-0.04 as to the integral neutron flux above 0.1 MeV and 1.30+-0.12 as to the reaction rate of indium foil. Use were made of the one and two dimensional neutron transport code ANISN and TWOTRAN for evaluation by computation. The reaction rates calculated by infinite slab model with ANISN code agree well with the experiments when normalized at the source point where neutrons are incident on the grid plate shield. (auth.)

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

Thickness Optimisation of Textiles Subjected to Heat and Mass Transport during Ironing

Directory of Open Access Journals (Sweden)

Korycki Ryszard

2016-09-01

Full Text Available Let us next analyse the coupled problem during ironing of textiles, that is, the heat is transported with mass whereas the mass transport with heat is negligible. It is necessary to define both physical and mathematical models. Introducing two-phase system of mass sorption by fibres, the transport equations are introduced and accompanied by the set of boundary and initial conditions. Optimisation of material thickness during ironing is gradient oriented. The first-order sensitivity of an arbitrary objective functional is analysed and included in optimisation procedure. Numerical example is the thickness optimisation of different textile materials in ironing device.

Thermohydraulic and thermal stress aspects of a porous blockage in an LMFBR fuel assembly

International Nuclear Information System (INIS)

Kuzay, T.M.; Marr, W.W.; Helenberg, H.W.; Ariman, T.; Wilson, R.E.; Pedersen, D.R.

1979-01-01

The current safety scenarios of Liquid Metal Fast Breeder Reactors (LMFBR) under local fault propagation include the study of a hypothetical accident initiated by the formation of an external debris porous blockage in a fuel subassembly. In this preliminary experimental and analytical investigation, a non-heat-generating porous blockage was postulated to cover 18 flow channels of a 37 pin Fast Test Reactor (FTR) type fuel subassembly. The axial extent of the blockage is 50 mm. The blockage material is stainless steel (SS 316) with 30 percent average porosity (percent void volume). The blockage and the pins were modeled with a finite element technique and the thermal field in the blockage was predicted. This thermal field was utilized to do a planar thermal stress analysis of the postulated blockage. To verify the analytical model and also to better understand the thermal-hydraulics of such a porous blockage out-of-pile tests were conducted in a sodium loop. Data from the out-of-pile tests was utilized to calibrate and improve the analytical model

Systems with a constant heat flux with applications to radiative heat transport across nanoscale gaps and layers

Science.gov (United States)

Budaev, Bair V.; Bogy, David B.

2018-06-01

We extend the statistical analysis of equilibrium systems to systems with a constant heat flux. This extension leads to natural generalizations of Maxwell-Boltzmann's and Planck's equilibrium energy distributions to energy distributions of systems with a net heat flux. This development provides a long needed foundation for addressing problems of nanoscale heat transport by a systematic method based on a few fundamental principles. As an example, we consider the computation of the radiative heat flux between narrowly spaced half-spaces maintained at different temperatures.

Basic analysis and a comparison of the characteristics GCFRs and the LMFBR with the thorium cycle in one-group diffusion theory

International Nuclear Information System (INIS)

Sabundjian, G.; Ishiguro, Y.

1991-09-01

A preliminary study of neutronics of thorium cycle fast breeder reactor has been done using simplified reactor models and analyses methods with the aim of finding a type of breeder reactor suitable for an efficient utilization of thorium that is abundant in Brazil. Basic methods of cross section processing and reactor calculation are studied and applied to analyse breeding characteristics of GCFRs and LMFBRs. The GCFR is fueled with oxide pins and cooled with helium. The LMFBR is fueled with thin metallic pins to achieve high power densities. Neutronics characteristics are determined as functions of the average power density and the fuel volume fraction. Results show that a high power density and a high fuel volume fraction are desirable to achieve short doubling times, that the GCFR is inferior to the LMFBR in regard to the doubling time and that the LMFBR can achieve reactor doubling times ten years with an average power density of ∼ 600MW/m 3 and fuel volume fraction of 40%. (author)

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

Study of the electron heat transport in Tore-Supra tokamak

International Nuclear Information System (INIS)

Harauchamps, E.

2004-01-01

This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

Science.gov (United States)

Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

2014-01-07

A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

Science.gov (United States)

Kelly, Sean M.; Kromer, Brian R.; Litwin, Michael M.; Rosen, Lee J.; Christie, Gervase Maxwell; Wilson, Jamie R.; Kosowski, Lawrence W.; Robinson, Charles

2016-01-19

A method and apparatus for producing heat used in a synthesis gas production process is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the steam reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5

Analysis for Heat Transfer in a High Current-Passing Carbon Nanosphere Using Nontraditional Thermal Transport Model.

Science.gov (United States)

Hol C Y; Chen, B C; Tsai, Y H; Ma, C; Wen, M Y

2015-11-01

This paper investigates the thermal transport in hollow microscale and nanoscale spheres subject to electrical heat source using nontraditional thermal transport model. Working as supercapacitor electrodes, carbon hollow micrometer- and nanometer-sized spheres needs excellent heat transfer characteristics to maintain high specific capacitance, long cycle life, and high power density. In the nanoscale regime, the prediction of heat transfer from the traditional heat conduction equation based on Fourier's law deviates from the measured data. Consequently, the electrical heat source-induced heat transfer characteristics in hollow micrometer- and nanometer-sized spheres are studied using nontraditional thermal transport model. The effects of parameters on heat transfer in the hollow micrometer- and nanometer-sized spheres are discussed in this study. The results reveal that the heat transferred into the spherical interior, temperature and heat flux in the hollow sphere decrease with the increasing Knudsen number when the radius of sphere is comparable to the mean free path of heat carriers.

Numerical simulation of the transport phenomena due to sudden heating in porous media

Energy Technology Data Exchange (ETDEWEB)

Lei, S.Y.; Zheng, G.Y.; Wang, B.X.; Yang, R.G.; Xia, C.M.

1997-07-01

Such process as wet porous media suddenly heated by hot fluids frequently occurs in nature and in industrial applications. The three-variable simulation model was developed to predict violent transport phenomena due to sudden heating in porous media. Two sets of independent variables were applied to different regions in porous media in the simulation. For the wet zone, temperature, wet saturation and air pressure were used as the independent variables. For the dry zone, the independent variables were temperature, vapor pressure and air pressure. The model simulated two complicated transport processes in wet unsaturated porous media which is suddenly heated by melting metal or boiling water. The effect of the gas pressure is also investigated on the overall transport phenomena.

VS2DRTI: Simulating Heat and Reactive Solute Transport in Variably Saturated Porous Media.

Science.gov (United States)

Healy, Richard W; Haile, Sosina S; Parkhurst, David L; Charlton, Scott R

2018-01-29

Variably saturated groundwater flow, heat transport, and solute transport are important processes in environmental phenomena, such as the natural evolution of water chemistry of aquifers and streams, the storage of radioactive waste in a geologic repository, the contamination of water resources from acid-rock drainage, and the geologic sequestration of carbon dioxide. Up to now, our ability to simulate these processes simultaneously with fully coupled reactive transport models has been limited to complex and often difficult-to-use models. To address the need for a simple and easy-to-use model, the VS2DRTI software package has been developed for simulating water flow, heat transport, and reactive solute transport through variably saturated porous media. The underlying numerical model, VS2DRT, was created by coupling the flow and transport capabilities of the VS2DT and VS2DH models with the equilibrium and kinetic reaction capabilities of PhreeqcRM. Flow capabilities include two-dimensional, constant-density, variably saturated flow; transport capabilities include both heat and multicomponent solute transport; and the reaction capabilities are a complete implementation of geochemical reactions of PHREEQC. The graphical user interface includes a preprocessor for building simulations and a postprocessor for visual display of simulation results. To demonstrate the simulation of multiple processes, the model is applied to a hypothetical example of injection of heated waste water to an aquifer with temperature-dependent cation exchange. VS2DRTI is freely available public domain software. © 2018, National Ground Water Association.

Turbulent transport regimes and the scrape-off layer heat flux width

Science.gov (United States)

Myra, J. R.; D'Ippolito, D. A.; Russell, D. A.

2015-04-01

Understanding the responsible mechanisms and resulting scaling of the scrape-off layer (SOL) heat flux width is important for predicting viable operating regimes in future tokamaks and for seeking possible mitigation schemes. In this paper, we present a qualitative and conceptual framework for understanding various regimes of edge/SOL turbulence and the role of turbulent transport as the mechanism for establishing the SOL heat flux width. Relevant considerations include the type and spectral characteristics of underlying instabilities, the location of the gradient drive relative to the SOL, the nonlinear saturation mechanism, and the parallel heat transport regime. We find a heat flux width scaling with major radius R that is generally positive, consistent with the previous findings [Connor et al., Nucl. Fusion 39, 169 (1999)]. The possible relationship of turbulence mechanisms to the neoclassical orbit width or heuristic drift mechanism in core energy confinement regimes known as low (L) mode and high (H) mode is considered, together with implications for the future experiments.

Turbulent transport regimes and the scrape-off layer heat flux width

International Nuclear Information System (INIS)

Myra, J. R.; D'Ippolito, D. A.; Russell, D. A.

2015-01-01

Understanding the responsible mechanisms and resulting scaling of the scrape-off layer (SOL) heat flux width is important for predicting viable operating regimes in future tokamaks and for seeking possible mitigation schemes. In this paper, we present a qualitative and conceptual framework for understanding various regimes of edge/SOL turbulence and the role of turbulent transport as the mechanism for establishing the SOL heat flux width. Relevant considerations include the type and spectral characteristics of underlying instabilities, the location of the gradient drive relative to the SOL, the nonlinear saturation mechanism, and the parallel heat transport regime. We find a heat flux width scaling with major radius R that is generally positive, consistent with the previous findings [Connor et al., Nucl. Fusion 39, 169 (1999)]. The possible relationship of turbulence mechanisms to the neoclassical orbit width or heuristic drift mechanism in core energy confinement regimes known as low (L) mode and high (H) mode is considered, together with implications for the future experiments

Dynamic simulation of LMFBR systems

International Nuclear Information System (INIS)

Agrawal, A.K.; Khatib-Rahbar, M.

1980-01-01

This review article focuses on the dynamic analysis of liquid-metal-cooled fast breeder reactor systems in the context of protected transients. Following a brief discussion on various design and simulation approaches, a critical review of various models for in-reactor components, intermediate heat exchangers, heat transport systems and the steam generating system is presented. A brief discussion on choice of fuels as well as core and blanket system designs is also included. Numerical considerations for obtaining system-wide steady-state and transient solutions are discussed, and examples of various system transients are presented. Another area of major interest is verification of phenomenological models. Various steps involved in the code and model verification are briefly outlined. The review concludes by posing some further areas of interest in fast reactor dynamics and safety. (author)

Universal high-temperature heat treatment furnace for FBR mixed uranium and plutonium carbide fuel

International Nuclear Information System (INIS)

Handa, Muneo; Takahashi, Ichiro; Watanabe, Hitoshi

1978-10-01

A universal high-temperature heat treatment furnace for LMFBR advanced fuels was installed in Plutonium Fuel Laboratory, Oarai Research Establishment. Design, construction and performance of the apparatus are described. With the apparatus, heat treatment of the fuel under a controlled gas atmosphere and quenching of the fuel with blowing helium gas are possible. Equipment to measure impurity gas release of the fuel is also provided. Various plutonium enclosure techniques, e.g., a gas line filter with new exchange mechanics, have been developed. In performance test, results of the enclosure techniques are described. (author)

Optimizing the design of large-scale ground-coupled heat pump systems using groundwater and heat transport modeling

Energy Technology Data Exchange (ETDEWEB)

Fujii, H.; Itoi, R.; Fujii, J. [Kyushu University, Fukuoka (Japan). Faculty of Engineering, Department of Earth Resources Engineering; Uchida, Y. [Geological Survey of Japan, Tsukuba (Japan)

2005-06-01

In order to predict the long-term performance of large-scale ground-coupled heat pump (GCHP) systems, it is necessary to take into consideration well-to-well interference, especially in the presence of groundwater flow. A mass and heat transport model was developed to simulate the behavior of this type of system in the Akita Plain, northern Japan. The model was used to investigate different operational schemes and to maximize the heat extraction rate from the GCHP system. (author)

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

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.

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

Consequences of nonlinear heat transport laws on expected plasma profiles

International Nuclear Information System (INIS)

Lackner, K.

1987-03-01

The expected variation of plasma pressure profiles against changes in power deposition is investigated by using a simple linear heat transport law as well as a quadratic one. Applying the quadratic transport law it can be shown that the stiffening of the resulting profiles is sufficient to understand the experimentally measured phenomenon of 'profile consistence' without further assumptions of nonlocal effects. (orig.) [de

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

Study on a neon cryogenic oscillating heat pipe with long heat transport distance

Science.gov (United States)

Liang, Qing; Li, Yi; Wang, Qiuliang

2017-12-01

An experimental study is carried out to study the heat transfer characteristics of a cryogenic oscillating heat pipe (OHP) with long heat transport distance. The OHP is made up of a capillary tube with an inner diameter of 1.0 mm and an outer diameter of 2.0 mm. The working fluid is neon, and the length of the adiabatic section is 480 mm. Tests are performed with the different heat inputs, liquid filling ratios and condenser temperature. For the cryogenic OHP with a liquid filling ratio of 30.7% at the condenser temperature of 28 K, the effective thermal conductivity is 3466-30,854 W/m K, and the maximum transfer power is 35.60 W. With the increment of the heat input, the effective thermal conductivity of the cryogenic OHP increases at the liquid filling ratios of 30.7% and 38.5%, while it first increases and then decreases at the liquid filling ratios of 15.2% and 23.3%. Moreover, the effective thermal conductivity increases with decreasing liquid filling ratio at the small heat input, and the maximum transfer power first increases and then decreases with increasing liquid filling ratio. Finally, it is found that the thermal performance of the cryogenic OHP can be improved by increasing the condenser temperature.

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.

Diffusive heat transport across magnetic islands and stochastic layers in tokamaks

International Nuclear Information System (INIS)

Hoelzl, Matthias

2010-01-01

Heat transport in tokamak plasmas with magnetic islands and ergodic field lines was simulated at realistic plasma parameters in realistic tokamak geometries. This requires the treatment of anisotropic heat diffusion, which is more efficient along magnetic field lines by up to ten orders of magnitude than perpendicular to them. Comparisons with analytical predictions and experimental measurements allow to determine the stability properties of neoclassical tearing modes as well as the experimental heat diffusion anisotropy.

Heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

International Nuclear Information System (INIS)

Travis, J.R.

1985-01-01

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen diffusion flames in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes. 18 refs., 24 figs

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

Three dimensional heat transport modeling in Vossoroca reservoir

Science.gov (United States)

Arcie Polli, Bruna; Yoshioka Bernardo, Julio Werner; Hilgert, Stephan; Bleninger, Tobias

2017-04-01

Freshwater reservoirs are used for many purposes as hydropower generation, water supply and irrigation. In Brazil, according to the National Energy Balance of 2013, hydropower energy corresponds to 70.1% of the Brazilian demand. Superficial waters (which include rivers, lakes and reservoirs) are the most used source for drinking water supply - 56% of the municipalities use superficial waters as a source of water. The last two years have shown that the Brazilian water and electricity supply is highly vulnerable and that improved management is urgently needed. The construction of reservoirs affects physical, chemical and biological characteristics of the water body, e.g. stratification, temperature, residence time and turbulence reduction. Some water quality issues related to reservoirs are eutrophication, greenhouse gas emission to the atmosphere and dissolved oxygen depletion in the hypolimnion. The understanding of the physical processes in the water body is fundamental to reservoir management. Lakes and reservoirs may present a seasonal behavior and stratify due to hydrological and meteorological conditions, and especially its vertical distribution may be related to water quality. Stratification can control heat and dissolved substances transport. It has been also reported the importance of horizontal temperature gradients, e.g. inflows and its density and processes of mass transfer from shallow to deeper regions of the reservoir, that also may impact water quality. Three dimensional modeling of the heat transport in lakes and reservoirs is an important tool to the understanding and management of these systems. It is possible to estimate periods of large vertical temperature gradients, inhibiting vertical transport and horizontal gradients, which could be responsible for horizontal transport of heat and substances (e.g. differential cooling or inflows). Vossoroca reservoir was constructed in 1949 by the impoundment of São João River and is located near to

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

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.

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

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.

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

Momentum, heat, and mass transfer analogy for vertical hydraulic transport of inert particles

Directory of Open Access Journals (Sweden)

Jaćimovski Darko R.

2014-01-01

Full Text Available Wall-to-bed momentum, heat and mass transfer in vertical liquid-solids flow, as well as in single phase flow, were studied. The aim of this investigation was to establish the analogy among those phenomena. Also, effect of particles concentration on momentum, heat and mass transfer was studied. The experiments in hydraulic transport were performed in a 25.4 mm I.D. cooper tube equipped with a steam jacket, using spherical glass particles of 1.94 mm in diameter and water as a transport fluid. The segment of the transport tube used for mass transfer measurements was inside coated with benzoic acid. In the hydraulic transport two characteristic flow regimes were observed: turbulent and parallel particle flow regime. The transition between two characteristic regimes (γ*=0, occurs at a critical voidage ε≈0.85. The vertical two-phase flow was considered as the pseudofluid, and modified mixture-wall friction coefficient (fw and modified mixture Reynolds number (Rem were introduced for explanation of this system. Experimental data show that the wall-to-bed momentum, heat and mass transfer coefficients, in vertical flow of pseudofluid, for the turbulent regime are significantly higher than in parallel regime. Wall-to-bed, mass and heat transfer coefficients in hydraulic transport of particles were much higher then in single-phase flow for lower Reynolds numbers (Re15000, there was not significant difference. The experimental data for wall-to-bed momentum, heat and mass transfer in vertical flow of pseudofluid in parallel particle flow regime, show existing analogy among these three phenomena. [Projekat Ministarstva nauke Republike Srbije, br. 172022

Heat Transport in Graphene Ferromagnet-Insulator-Superconductor Junctions

Institute of Scientific and Technical Information of China (English)

LI Xiao-Wei

2011-01-01

We study heat transport in a graphene ferromagnet-insulator-superconducting junction. It is found that the thermal conductance of the graphene ferromagnet-insulator-superconductor (FIS) junction is an oscillatory function of the barrier strength x in the thin-barrier limit. The gate potential U0 decreases the amplitude of thermal conductance oscillation. Both the amplitude and phase of the thermal conductance oscillation varies with the exchange energy Eh. The thermal conductance of a graphene FIS junction displays the usual exponential dependence on temperature, reflecting the s-wave symmetry of superconducting graphene.%@@ We study heat transport in a graphene ferromagnet-insulator-superconducting junction.It is found that the thermal conductance of the graphene ferromagnet-insulator-superconductor(FIS)junction is an oscillatory function of the barrier strength X in the thin-barrier limit.The gate potential Uo decreases the amplitude of thermal conductance oscillation.Both the amplitude and phase of the thermal conductance oscillation varies with the exchange energy Eh.The thermal conductance of a graphene FIS junction displays the usual exponential dependence on temperature, reflecting the s-wave symmetry of superconducting graphene.

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.)

Poloidal profiles and transport during turbulent heating

International Nuclear Information System (INIS)

Mascheroni, P.L.

1977-01-01

The current penetration stage of a turbulently heated tokamak is modeled. The basic formulae are written in slab geometry since the dominant anomalous transport has a characteristic frequency much larger than the bounce frequency. Thus, the basic framework is provided by the Maxwell and fluid equations, with classical and anomalous transport. Quasi-neutrality is used. It is shown that the anomalous collision frequency dominates the anomalous viscosity and thermal conductivity, and that the convective wave transport can be neglected. For these numerical estimates, the leading term in the quasi-linear series is used. During the current penetration stage the distribution function for the particles will depart from a single Maxwellian type. Hence, the first objective was to numerically compare calculated poloidal magnetic field profiles with measured, published poloidal profiles. The poloidal magnetic field has been calculated using a code which handles the anomalous collision frequency self-consistently. The agreement is good, and it is concluded that the current penetration stage can be satisfactorily described by this model

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)

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.)

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

TETIG diagrams - a new way to optimise the design parameters and heat treatment of joints made in high-temperature brazing alloys. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Johnson, R. (UKAEA Springfields Nuclear Power Development Labs.)

1982-12-01

The applications and problems of brazing are reviewed. Phase studies with the braze filler metal chosen for the LMFBR 9% Cr 1% Mo tube-in-tube joint work (BNi4), are discussed, with special reference to the problem of how to eliminate the centre-line eutectics containing hard, brittle compounds. A TETIG diagram is explained with reference to the variables (1) temperature of brazing operation; (2) time of soaking at temperature; and (3) the gap within the joints. Experiments are reported on brazing specimens of AISI 321 stainless steel, using braze filler metals containing various proportions of boron and silicon as the melting point temperature depressant. TETIG diagrams are constructed and used to predict how to optimize further joints. Micrographs show the effects of the variables on the microstructures.

Mobile heat storage containers and their transport by rail or road

Energy Technology Data Exchange (ETDEWEB)

Goldenberg, Philipp

2013-10-15

Mobile heat storage containers are capable of making a contribution to the meaningful use of energy which is needed for use at a location other than where it originates. The study presented in this report outlines the technology of mobile heat storage and analyses an example of its transport by rail or road. (orig.)

Process for the transport of heat energy released by a nuclear reactor

International Nuclear Information System (INIS)

Nuernberg, H.W.; Wolff, G.

1978-01-01

The heat produced in a nuclear reactor is converted into latent chemical binding energy. The heat can be released again below 400 0 C by recombination after transport by decomposition of ethane or propane into ethylene or propylene and hydrogen. (TK) [de

Studies of heat transport to forced-flow He II

International Nuclear Information System (INIS)

Dresner, L.; Kashani, A.; Van Sciver, S.W.

1985-01-01

Analytical and experimental studies of heat transport to forced-flow He II are reported. The work is pertinent to the transfer of He II in space. An analytical model has been developed that establishes a condition for two-phase flow to occur in the transfer line. This condition sets an allowable limit to the heat leak into the transfer line. Experimental measurements of pressure drop and flow meter performances indicate that turbulent He II can be analyzed in terms of classical pressure drop correlations

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

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

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)

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

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)

Required momentum, heat, and mass transport experiments for liquid-metal blankets

International Nuclear Information System (INIS)

Tillack, M.S.; Sze, D.K.; Abdou, M.A.

1986-01-01

Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Energy Technology Data Exchange (ETDEWEB)

Wolf, S.; Holmes, D.H.

1977-04-01

An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

Heat transport in the quasi-single-helicity islands of EXTRAP T2R

Science.gov (United States)

Frassinetti, L.; Brunsell, P. R.; Drake, J.

2009-03-01

The heat transport inside the magnetic island generated in a quasi-single-helicity regime of a reversed-field pinch device is studied by using a numerical code that simulates the electron temperature and the soft x-ray emissivity. The heat diffusivity χe inside the island is determined by matching the simulated signals with the experimental ones. Inside the island, χe turns out to be from one to two orders of magnitude lower than the diffusivity in the surrounding plasma, where the magnetic field is stochastic. Furthermore, the heat transport properties inside the island are studied in correlation with the plasma current and with the amplitude of the magnetic fluctuations.

Characteristics of nonlocally-coupled transition of the heat transport in LHD

International Nuclear Information System (INIS)

Tamura, N.; Ida, K.; Tanaka, K.; Tokuzawa, T.; Itoh, K.; Shimozuma, T.; Kubo, S.; Tsuchiya, H.; Nagayama, Y.; Kawahata, K.; Sudo, S.; Yamada, H.; Inagaki, S.

2010-01-01

A comparison of characteristics between a nonlocal transport phenomenon and an electron internal transport barrier (ITB) in the Large Helical Device is performed with a transient transport analysis and from the viewpoint of a dynamic behavior of transport state. The electron ITB is characterized by a jump of electron temperature gradient. In contrast, the transient transport analysis indicates the nonlocal transport phenomenon is characterized by a jump of electron heat flux. And seen from the viewpoint of the dynamic behavior of transport state, the physical mechanism of the appearance of the nonlocal transport phenomenon is found to be qualitatively different from that of the formation of the electron ITB. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Monju secondary heat transport system sodium leak

International Nuclear Information System (INIS)

Suzuki, Takeo; Hiroi, Hiroshi; Usami, Shin; Iwata, Koji.

1996-01-01

On December 8, 1995, the sodium leakage from the secondary heat transport system (SHTS) occurred in the piping room of the reactor auxiliary building in Monju. The secondary sodium leaked through a temperature sensor, due to the breakaway of the tip of the well tube of the sensor installed near the outlet of the intermediate heat exchanger (IHX) in the C loop of SHTS. The reactor core remained cooled and thus, from the viewpoint of radiological hazards, the safety of the reactor was secured. There were no adverse effects for operating personnel or the surrounding environment. The cause of the well tube failure is considered to result from high cycle fatigue due to flow induced vibrations. Delay in draining the sodium from the leaking loop increased the consequential effects from sodium combustion products. (author)

EFFECT OF SANDSTONE ANISOTROPY ON ITS HEAT AND MOISTURE TRANSPORT PROPERTIES

Directory of Open Access Journals (Sweden)

Jan Fořt

2015-09-01

Full Text Available Each type of natural stone has its own geological history, formation conditions, different chemical and mineralogical composition, which influence its possible anisotropy. Knowledge in the natural stones anisotropy represents crucial information for the process of stone quarrying, its correct usage and arrangement in building applications. Because of anisotropy, many natural stones exhibit different heat and moisture transport properties in various directions. The main goal of this study is to analyse several anisotropy indices and their effect on heat transport and capillary absorption. For the experimental determination of the anisotropy effect, five types of sandstone coming from different operating quarries in the Czech Republic are chosen. These materials are often used for restoration of culture heritage monuments as well as for other building applications where they are used as facing slabs, facade panels, decoration stones, paving, etc. For basic characterization of studied materials, determination of their bulk density, matrix density and total open porosity is done. Chemical composition of particular sandstones is analysed by X-Ray Fluorescence. Anisotropy is examined by the non-destructive measurement of velocity of ultrasonic wave propagation. On the basis of ultrasound testing data, the relative anisotropy, total anisotropy and anisotropy coefficient are calculated. Then, the measurement of thermal conductivity and thermal diffusivity in various directions of samples orientation is carried out. The obtained results reveal significant differences between the parameters characterizing the heat transport in various directions, whereas these values are in accordance with the indices of anisotropy. Capillary water transport is described by water absorption coefficient measured using a sorption experiment, which is performed for distilled water and 1M NaCl water solution.  The measured data confirm the effect of anisotropy which is

The heat and moisture transport properties of wet porous media

International Nuclear Information System (INIS)

Wang, B.X.; Fang, Z.H.; Yu, W.P.

1989-01-01

Existing methods for determining heat and moisture transport properties in porous media are briefly reviewed, and their merits and deficiencies are discussed. Emphasis is placed on research in developing new transient methods undertaken in China during the recent years. An attempt has been made to relate the coefficients in the heat and mass transfer equations with inherent properties of the liquid and matrix and then to predict these coefficients based on limited measurements

1D momentum-conserving systems: the conundrum of anomalous versus normal heat transport

International Nuclear Information System (INIS)

Li, Yunyun; Li, Nianbei; Hänggi, Peter; Li, Baowen; Liu, Sha

2015-01-01

Transport and the spread of heat in Hamiltonian one dimensional momentum conserving nonlinear systems is commonly thought to proceed anomalously. Notable exceptions, however, do exist of which the coupled rotator model is a prominent case. Therefore, the quest arises to identify the origin of manifest anomalous energy and momentum transport in those low dimensional systems. We develop the theory for both, the statistical densities for momentum- and energy-spread and particularly its momentum-/heat-diffusion behavior, as well as its corresponding momentum/heat transport features. We demonstrate that the second temporal derivative of the mean squared deviation of the momentum spread is proportional to the equilibrium correlation of the total momentum flux. Subtracting the part which corresponds to a ballistic momentum spread relates (via this integrated, subleading momentum flux correlation) to an effective viscosity, or equivalently, to the underlying momentum diffusivity. We next put forward the intriguing hypothesis: normal spread of this so adjusted excess momentum density causes normal energy spread and alike normal heat transport (Fourier Law). Its corollary being that an anomalous, superdiffusive broadening of this adjusted excess momentum density in turn implies an anomalous energy spread and correspondingly anomalous, superdiffusive heat transport. This hypothesis is successfully corroborated within extensive molecular dynamics simulations over large extended time scales. Our numerical validation of the hypothesis involves four distinct archetype classes of nonlinear pair-interaction potentials: (i) a globally bounded pair interaction (the noted coupled rotator model), (ii) unbounded interactions acting at large distances (the coupled rotator model amended with harmonic pair interactions), (iii) the case of a hard point gas with unbounded square-well interactions and (iv) a pair interaction potential being unbounded at short distances while displaying an

1D momentum-conserving systems: the conundrum of anomalous versus normal heat transport

Science.gov (United States)

Li, Yunyun; Liu, Sha; Li, Nianbei; Hänggi, Peter; Li, Baowen

2015-04-01

Transport and the spread of heat in Hamiltonian one dimensional momentum conserving nonlinear systems is commonly thought to proceed anomalously. Notable exceptions, however, do exist of which the coupled rotator model is a prominent case. Therefore, the quest arises to identify the origin of manifest anomalous energy and momentum transport in those low dimensional systems. We develop the theory for both, the statistical densities for momentum- and energy-spread and particularly its momentum-/heat-diffusion behavior, as well as its corresponding momentum/heat transport features. We demonstrate that the second temporal derivative of the mean squared deviation of the momentum spread is proportional to the equilibrium correlation of the total momentum flux. Subtracting the part which corresponds to a ballistic momentum spread relates (via this integrated, subleading momentum flux correlation) to an effective viscosity, or equivalently, to the underlying momentum diffusivity. We next put forward the intriguing hypothesis: normal spread of this so adjusted excess momentum density causes normal energy spread and alike normal heat transport (Fourier Law). Its corollary being that an anomalous, superdiffusive broadening of this adjusted excess momentum density in turn implies an anomalous energy spread and correspondingly anomalous, superdiffusive heat transport. This hypothesis is successfully corroborated within extensive molecular dynamics simulations over large extended time scales. Our numerical validation of the hypothesis involves four distinct archetype classes of nonlinear pair-interaction potentials: (i) a globally bounded pair interaction (the noted coupled rotator model), (ii) unbounded interactions acting at large distances (the coupled rotator model amended with harmonic pair interactions), (iii) the case of a hard point gas with unbounded square-well interactions and (iv) a pair interaction potential being unbounded at short distances while displaying an

Heat transport in the XXZ spin chain: from ballistic to diffusive regimes and dephasing enhancement

International Nuclear Information System (INIS)

Mendoza-Arenas, J J; Al-Assam, S; Clark, S R; Jaksch, D

2013-01-01

In this work we study the heat transport in an XXZ spin-1/2 Heisenberg chain with homogeneous magnetic field, incoherently driven out of equilibrium by reservoirs at the boundaries. We focus on the effect of bulk dephasing (energy-dissipative) processes in different parameter regimes of the system. The non-equilibrium steady state of the chain is obtained by simulating its evolution under the corresponding Lindblad master equation, using the time evolving block decimation method. In the absence of dephasing, the heat transport is ballistic for weak interactions, while being diffusive in the strongly interacting regime, as evidenced by the heat current scaling with the system size. When bulk dephasing takes place in the system, diffusive transport is induced in the weakly interacting regime, with the heat current monotonically decreasing with the dephasing rate. In contrast, in the strongly interacting regime, the heat current can be significantly enhanced by dephasing for systems of small size. (paper)

A practical nonlocal model for heat transport in magnetized laser plasmas

International Nuclear Information System (INIS)

Nicolaie, Ph.D.; Feugeas, J.-L.A.; Schurtz, G.P.

2006-01-01

A model of nonlocal transport for multidimensional radiation magnetohydrodynamics codes is presented. In laser produced plasmas, it is now believed that the heat transport can be strongly modified by the nonlocal nature of the electron conduction. Other mechanisms, such as self-generated magnetic fields, may also affect the heat transport. The model described in this work, based on simplified Fokker-Planck equations aims at extending the model of G. Schurtz, Ph. Nicolaie, and M. Busquet [Phys. Plasmas 7, 4238 (2000)] to magnetized plasmas. A complete system of nonlocal equations is derived from kinetic equations with self-consistent electric and magnetic fields. These equations are analyzed and simplified in order to be implemented into large laser fusion codes and coupled to other relevant physics. The model is applied to two laser configurations that demonstrate the main features of the model and point out the nonlocal Righi-Leduc effect in a multidimensional case

Controlling Heat Transport and Flow Structures in Thermal Turbulence Using Ratchet Surfaces

Science.gov (United States)

Jiang, Hechuan; Zhu, Xiaojue; Mathai, Varghese; Verzicco, Roberto; Lohse, Detlef; Sun, Chao

2018-01-01

In this combined experimental and numerical study on thermally driven turbulence in a rectangular cell, the global heat transport and the coherent flow structures are controlled with an asymmetric ratchetlike roughness on the top and bottom plates. We show that, by means of symmetry breaking due to the presence of the ratchet structures on the conducting plates, the orientation of the large scale circulation roll (LSCR) can be locked to a preferred direction even when the cell is perfectly leveled out. By introducing a small tilt to the system, we show that the LSCR orientation can be tuned and controlled. The two different orientations of LSCR give two quite different heat transport efficiencies, indicating that heat transport is sensitive to the LSCR direction over the asymmetric roughness structure. Through a quantitative analysis of the dynamics of thermal plume emissions and the orientation of the LSCR over the asymmetric structure, we provide a physical explanation for these findings. The current work has important implications for passive and active flow control in engineering, biofluid dynamics, and geophysical flows.

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.)

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

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

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

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

Heat transport modelling in EXTRAP T2R

Science.gov (United States)

Frassinetti, L.; Brunsell, P. R.; Cecconello, M.; Drake, J. R.

2009-02-01

A model to estimate the heat transport in the EXTRAP T2R reversed field pinch (RFP) is described. The model, based on experimental and theoretical results, divides the RFP electron heat diffusivity χe into three regions, one in the plasma core, where χe is assumed to be determined by the tearing modes, one located around the reversal radius, where χe is assumed not dependent on the magnetic fluctuations and one in the extreme edge, where high χe is assumed. The absolute values of the core and of the reversal χe are determined by simulating the electron temperature and the soft x-ray and by comparing the simulated signals with the experimental ones. The model is used to estimate the heat diffusivity and the energy confinement time during the flat top of standard plasmas, of deep F plasmas and of plasmas obtained with the intelligent shell.

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory, August 1977

International Nuclear Information System (INIS)

Fee, G.G.

1977-01-01

Technical highlights are presented for the following safety-related studies: heavy section steel technology, fission product beta and gamma energy release, fission product release from LWR fuel, fission product transport tests, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, design criteria for piping and nozzles, and noise diagnostics for safety assessment

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)

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

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.

Flexibility analysis of main primary heat transport system : Narora Atomic Power Project

International Nuclear Information System (INIS)

Rastogi, S.K.

1975-01-01

The paper presents flexibility analysis problem of main primary heat transport system and the approximate analysis that has been made to estimate the loads coming on major equipments. The primary heat transport system for Narora Atomic Power Project is adopting vertical steam generators and pumps equally divided on either side of the reactor with inter-connecting pipes and feeders. Since the system is mainly spring supported with movement of a few points in certain direction defined but no anchorage, it represents a good problem for flexibility analysis which can only be solved in one step by developing a good computer programme. (author)

A continuum self organized critically model of turbulent heat transport in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Tangri, V; Das, A; Kaw, P; Singh, R [Institute for Plasma Research, Gandhinagar (India)

2003-09-01

Based on the now well known and experimentally observed critical gradient length (R/L{sub Te} = RT/{nabla}T) in tokamaks, we present a continuum one dimensional model for explaining self organized heat transport in tokamaks. Key parameters of this model include a novel hysteresis parameter which ensures that the switch of heat transport coefficient {chi} upwards and downwards takes place at two different values of R/L{sub Te}. Extensive numerical simulations of this model reproduce many features of present day tokamaks such as submarginal temperature profiles, intermittent transport events, 1/f scaling of the frequency spectra, propagating fronts, etc. This model utilises a minimal set of phenomenological parameters, which may be determined from experiments and/or simulations. Analytical and physical understanding of the observed features has also been attempted. (author)

Heating and active control of profiles and transport by IBW in the HT-7 tokamak

International Nuclear Information System (INIS)

Zhao Yanping; Wan Baonian; Li Jiangang

2003-01-01

Significant progress on Ion Bernstein Wave (IBW) heating and control of profiles has been obtained in HT-7. Both on-axis and off-axis electron heating with global peaked and local steep electron pressure profiles were realized if the position of the resonant layer was selected to be plasma far from the plasma edge region. Reduction of electron heat transport has been observed from sawtooth heat pulse propagation. Improvement of both particle and energy confinement was slight in the on-axis and considerable in the off-axis heating cases. The improved confinement in off-axis heating mode may be due to the extension of the high performance plasma volume caused by IBW. These studies demonstrate that IBWs are potentially a tool for active control of plasma profiles and transport. (author)

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)

A practical nonlocal model for heat transport in magnetized laser plasmas

Science.gov (United States)

Nicolaï, Ph. D.; Feugeas, J.-L. A.; Schurtz, G. P.

2006-03-01

A model of nonlocal transport for multidimensional radiation magnetohydrodynamics codes is presented. In laser produced plasmas, it is now believed that the heat transport can be strongly modified by the nonlocal nature of the electron conduction. Other mechanisms, such as self-generated magnetic fields, may also affect the heat transport. The model described in this work, based on simplified Fokker-Planck equations aims at extending the model of G. Schurtz, Ph. Nicolaï, and M. Busquet [Phys. Plasmas 7, 4238 (2000)] to magnetized plasmas. A complete system of nonlocal equations is derived from kinetic equations with self-consistent electric and magnetic fields. These equations are analyzed and simplified in order to be implemented into large laser fusion codes and coupled to other relevant physics. The model is applied to two laser configurations that demonstrate the main features of the model and point out the nonlocal Righi-Leduc effect in a multidimensional case.

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.)

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)

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)

Heating and transport in TFTR D-T plasmas

International Nuclear Information System (INIS)

Zarnstorff, M.C.; Scott, S.D.

1994-01-01

The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in TFTR. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. The radial profiles of the deuterium and tritium densities are determined from the DT fusion neutron emission profile

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

The US Nuclear Regulatory Commission aerosol release and the transport program

Energy Technology Data Exchange (ETDEWEB)

Silberberg, M; Kress, T [Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC (United States); Gieseke, J [Batelle Memorial Institute, Columbus, OH (United States)

1977-01-01

An overview is presented of the U.S.N.R.C. research program for providing experimentally verified, quantitative methods for estimating the release and transport of sodium and radionuclide aerosols following postulated accidents. The program is directed towards radiological consequence assessment, however a number of aerosol behavior mechanisms being studied are applicable to LMFBR operational considerations. Related theoretical and experimental work on aerosol formation, agglomeration, settling and plating is noted. (author)

Heating and active control of profiles and transport by IBW in the HT-7 tokamak

International Nuclear Information System (INIS)

Zhao Yanping

2002-01-01

By a series of technical improvements and intensive RF boronization, significant progresses on the IBW heating and control of profiles and transport has been obtained since last IAEA meeting. Both on-axis and off-axis electron heating with global peaked and local steeped electron pressure profile was realized if the resonant layer is in plasma far from the edge region. Maximum increment of electron temperature was about 2 keV at power of 200 kW. The heating factor reached 9.4 eV x 10 13 cm -3 /kW. Reduction of local electron heat transport around resonant layer has been observed. Significant improvement of particle confinement by a factor of 2-4 with very peaked density profile was obtained if 5/2-deuterium resonant layer is located at the plasma edge. Global transport and edge poloidal velocity shear can been controlled by IBW. (author)

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.

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

Thermal performance and heat transport in aquifer thermal energy storage

Science.gov (United States)

Sommer, W. T.; Doornenbal, P. J.; Drijver, B. C.; van Gaans, P. F. M.; Leusbrock, I.; Grotenhuis, J. T. C.; Rijnaarts, H. H. M.

2014-01-01

Aquifer thermal energy storage (ATES) is used for seasonal storage of large quantities of thermal energy. Due to the increasing demand for sustainable energy, the number of ATES systems has increased rapidly, which has raised questions on the effect of ATES systems on their surroundings as well as their thermal performance. Furthermore, the increasing density of systems generates concern regarding thermal interference between the wells of one system and between neighboring systems. An assessment is made of (1) the thermal storage performance, and (2) the heat transport around the wells of an existing ATES system in the Netherlands. Reconstruction of flow rates and injection and extraction temperatures from hourly logs of operational data from 2005 to 2012 show that the average thermal recovery is 82 % for cold storage and 68 % for heat storage. Subsurface heat transport is monitored using distributed temperature sensing. Although the measurements reveal unequal distribution of flow rate over different parts of the well screen and preferential flow due to aquifer heterogeneity, sufficient well spacing has avoided thermal interference. However, oversizing of well spacing may limit the number of systems that can be realized in an area and lower the potential of ATES.

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)

A simulation of heat transfer during billet transport

Energy Technology Data Exchange (ETDEWEB)

Jaklic, A.; Glogovac, B. [Institute of Metals and Technology, Ljubljana (Slovenia); Kolenko, T. [University of Ljubljana (Slovenia). Faculty of Natural Science and Technology; Zupancic, B. [University of Ljubljana (Slovenia). Faculty of Electrical Engineering; Zak, B. T. [Terming d.o.o., Ljubljana (Slovenia)

2002-07-01

This paper presents a simulation model for billet cooling during the billet's transport from the reheating furnace to the rolling mill. During the transport, the billet is exposed to radiation, convection and conduction. Due to the rectangular shape of the billet, the three-dimensional finite-difference model could be applied to calculate the heat conduction inside the billet. The billets are reheated in a gas-fired walking-beam furnace and are exposed to scaling. The model takes into account the effect of the thin oxide scale. We proved that the scale significantly affects the temperature distribution in the billet and should not be neglected. The model was verified by using a thermal camera. (author)

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)

Comparison of transient electron heat transport in LHD helical and JT-60U tokamak plasmas

International Nuclear Information System (INIS)

Inagaki, S.; Ida, K.; Tamura, N.; Shimozuma, T.; Kubo, S.; Nagayama, Y.; Kawahata, K.; Sudo, S.; Ohkubo, K.; Takenaga, H.; Isayama, A.; Takizuka, T.; Kamada, Y.; Miura, Y.

2005-01-01

Transient transport experiments are performed in plasmas with and without Internal Transport Barrier (ITB) on LHD and JT-60U. The dependence of χ e on electron temperature, T e , and electron temperature gradient, ∇T e , is analyzed by an empirical non-linear heat transport model. In plasmas without ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation. The χ e depends on T e and ∇T e in JT-60U, while the ∇T e dependence is weak in LHD. Inside the ITB region, there is no or weak ∇T e dependence both in LHD and JT-60U. A cold pulse growing driven by the negative T e dependence of χ e is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U). (author)

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

The influence of meridional ice transport on Europa's ocean stratification and heat content

Science.gov (United States)

Zhu, P.; Manucharyan, G.; Thompson, A. F.; Goodman, J. C.; Vance, S.

2017-12-01

Jupiter's moon Europa likely hosts a saltwater ocean beneath its icy surface. Geothermal heating and rotating convection in the ocean may drive a global overturning circulation that redistributes heat vertically and meridionally, preferentially warming the ice shell at the equator. Here we assess thepreviously unconstrained influence of ocean-ice coupling on Europa's ocean stratification and heat transport. We demonstrate that a relatively fresh layer can form at the ice-ocean interface due to a meridional ice transport forced by the differential ice shell heating between the equator and the poles. We provide analytical and numerical solutions for the layer's characteristics, highlighting their sensitivity to critical ocean parameters. For a weakly turbulent and highly saline ocean, a strong buoyancy gradient at the base of the freshwater layer can suppress vertical tracer exchange with the deeper ocean. As a result, the freshwater layer permits relatively warm deep ocean temperatures.

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)

Controlling heat transport and flow structures in thermal turbulence using ratchet surfaces

Science.gov (United States)

Sun, Chao; Jiang, Hechuan; Zhu, Xiaojue; Mathai, Varghese; Verzicco, Roberto; Lohse, Detlef

2017-11-01

In this combined experimental and numerical study on thermally driven turbulence in a rectangular cell, the global heat transport and the coherent flow structures are controlled with an asymmetric ratchet-like roughness on the top and bottom plates. We show that, by means of symmetry breaking due to the presence of the ratchet structures on the conducting plates, the orientation of the Large Scale Circulation Roll (LSCR) can be locked to a preferred direction even when the cell is perfectly leveled out. By introducing a small tilt to the system, we show that the LSCR orientation can be tuned and controlled. The two different orientations of LSCR give two quite different heat transport efficiencies, indicating that heat transport is sensitive to the LSCR direction over the asymmetric roughness structure. Through analysis of the dynamics of thermal plume emissions and the orientation of the LSCR over the asymmetric structure, we provide a physical explanation for these findings. This work is financially supported by the Natural Science Foundation of China under Grant No. 11672156, the Dutch Foundation for Fundamental Research on Matter (FOM), the Dutch Technology Foundation (STW) and a VIDI Grant.

Heat transport as torsional responses and Keldysh formalism in a curved spacetime

OpenAIRE

Shitade, Atsuo

2013-01-01

We revisit a theory of heat transport in the light of a gauge theory of gravity and find the proper heat current with a corresponding gauge field, which yields the natural definitions of the heat magnetization and the Kubo-formula contribution to the thermal conductivity as torsional responses. We also develop a general framework for calculating gravitational responses by combining the Keldysh and Cartan formalisms. By using this framework, we explicitly calculate these two quantities and rep...

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)

Design considerations for CRBRP heat transport system piping operating at elevated temperatures

International Nuclear Information System (INIS)

Pollono, L.P.; Mello, R.M.

1979-01-01

The heat transport system sodium piping for the Clinch River Breeder Reactor Plant (CRBRP) within the reactor containment building must withstand high temperatures for long periods of time. Each phase of the mechanical design process of the piping system is influenced by elevated temperature considerations which include material thermal creep effects, ratchetting caused by rapid temperature transients and stress relaxation, and material degradation effects. The structural design philosophy taken to design the CRBRP piping operating in a high temperature environment is described. The resulting design of the heat transport system piping is presented along with a discussion of special features that resulted from the elevated temperature considerations

Ductile fracture behaviour of primary heat transport piping material ...

Indian Academy of Sciences (India)

R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

Abstract. Design of primary heat transport (PHT) piping of pressurised heavy water reactors (PHWR) has to ensure implementation of leak-before-break con- cepts. In order to be able to do so, the ductile fracture characteristics of PHT piping material have to be quantified. In this paper, the fracture resistance of SA333, Grade.

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

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.

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

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.)

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)

Thermal transport in low dimensions from statistical physics to nanoscale heat transfer

CERN Document Server

2016-01-01

Understanding non-equilibrium properties of classical and quantum many-particle systems is one of the goals of contemporary statistical mechanics. Besides its own interest for the theoretical foundations of irreversible thermodynamics(e.g. of the Fourier's law of heat conduction), this topic is also relevant to develop innovative ideas for nanoscale thermal management with possible future applications to nanotechnologies and effective energetic resources. The first part of the volume (Chapters 1-6) describes the basic models, the phenomenology and the various theoretical approaches to understand heat transport in low-dimensional lattices (1D e 2D). The methods described will include equilibrium and nonequilibrium molecular dynamics simulations, hydrodynamic and kinetic approaches and the solution of stochastic models. The second part (Chapters 7-10) deals with applications to nano and microscale heat transfer, as for instance phononic transport in carbon-based nanomaterials, including the prominent case of na...

Overview of U.S. LMFBR structural materials mechanical properties program

International Nuclear Information System (INIS)

Horak, J.A.; Purdy, C.M.

This paper presents the objective, scope, and status of the U.S. Department of Energy's Materials and Structures Program to develop a data base on mechanical properties of structural materials for out-of-core structures and components for LMFBRs. Information on the development of a reference data base on materials for the reactor system, reactor enclosure system, primary heat transport system, intermediate heat transport system, and steam generator system is included. In addition, the development of the data and analyses to account for the effects of temperature and stress, as well as water/steam, sodium, and radiation environments, is described. Plans for the development of alternative materials for future out-of-core applications are presented. (author)

Nonlocal heat transport and improved target design for x-ray heating studies at x-ray free electron lasers

Science.gov (United States)

Hoidn, Oliver; Seidler, Gerald T.

2018-01-01

The extremely high-power densities and short durations of single pulses of x-ray free electron lasers (XFELs) have opened new opportunities in atomic physics, where complex excitation-relaxation chains allow for high ionization states in atomic and molecular systems, and in dense plasma physics, where XFEL heating of solid-density targets can create unique dense states of matter having temperatures on the order of the Fermi energy. We focus here on the latter phenomena, with special emphasis on the problem of optimum target design to achieve high x-ray heating into the warm dense matter (WDM) state. We report fully three-dimensional simulations of the incident x-ray pulse and the resulting multielectron relaxation cascade to model the spatial energy density deposition in multicomponent targets, with particular focus on the effects of nonlocal heat transport due to the motion of high energy photoelectrons and Auger electrons. We find that nanoscale high-Z /low-Z multicomponent targets can give much improved energy density deposition in lower-Z materials, with enhancements reaching a factor of 100. This has three important benefits. First, it greatly enlarges the thermodynamic parameter space in XFEL x-ray heating studies of lower-Z materials. Second, it allows the use of higher probe photon energies, enabling higher-information content x-ray diffraction (XRD) measurements such as in two-color XFEL operations. Third, while this is merely one step toward optimization of x-ray heating target design, the demonstration of the importance of nonlocal heat transport establishes important common ground between XFEL-based x-ray heating studies and more traditional laser plasma methods.

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

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)

Internal fuel motion as an inherent shutdown mechanism for LMFBR accidents: PINEX-3, PINEX-2, and HUT 5-2A experiments

International Nuclear Information System (INIS)

Ferrell, P.C.; Porten, D.R.; Martin, F.J.

1981-01-01

The PINEX-2 experiment verified the concept of axial internal molten fuel motion within annular fuel, representing an inherent shutdown mechanism for hypothetical transient overpower excursions on the order of 5$/s. The PINEX-3 experiment, simulating a 50 cents/s transient overpower, showed that limitations on the effectiveness of fuel motion may arise from freezing of the fuel and blockage of the internal movement. Analysis of these experiments was performed to assess the physical processes that dominate fuel relocation potential and to apply them to prototypic LMFBR pin conditions. Results indicate that internal fuel motion should be reliable as a shutdown mechanism in LMFBR's for a range of reactivity insertion rates beyond presently available experimental data

Heat and momentum transport of ion internal transport barrier plasmas on Large Helical Device

International Nuclear Information System (INIS)

Nagaoka, K.; Ida, K.; Yoshinuma, M.

2010-11-01

The peaked ion-temperature profile with steep gradient so called ion internal transport barrier (ion ITB) was formed in the neutral beam heated plasmas on the Large Helical Device (LHD) and the high-ion-temperature regime of helical plasmas has been significantly extended. The ion thermal diffusivity in the ion ITB plasma decreases down to the neoclassical transport level. The heavy ion beam probe (HIBP) observed the smooth potential profile with negative radial electric field (ion root) in the core region where the ion thermal diffusivity decreases significantly. The large toroidal rotation was also observed in the ion ITB core and the transport of toroidal momentum was analyzed qualitatively. The decrease of momentum diffusivity with ion temperature increase was observed in the ion ITB core. The toroidal rotation driven by ion temperature gradient so called intrinsic rotation is also identified. (author)

A new treatment of the heat transport equation with a transport barrier and applications to ECRH experiments in Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Zou, X.L.; Giruzzi, A.G.; Bouquey, F.; Clary, J.; Darbos, C.; Lennholm, M.; Magne, R.; Segui, J.L. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee, 13 - Saint-Paul-lez-Durance (France); Clemencon, A. [MIT, Electrochemical Energy Laboratory, Cambridge, MA (United States); Guivarch, C. [Ecole Nationale des Ponts et Chaussees, 77 - Marne-la-Vallee (France)

2004-07-01

An exact analytical solution of the electron heat diffusion equation in a cylinder has been found with a step-like diffusion coefficient, plus a monomial increase in the radial direction and a constant damping term. This model is sufficiently general to describe heat diffusion in the presence of a critical gradient threshold or a transport barrier, superimposed to the usual trend of increasing heat diffusivity from the plasma core to the edge. This type of representation allows us to see some well-known properties of heat transport phenomena in a different light. For instance, it has been shown that the contributions of the Eigenmodes to the time dependent solution grow at speeds that depend on the Eigenmode order i.e. at the beginning of the heating phase all the Eigenmodes are equally involved, whereas at the end only the lower order ones are left. This implies, e.g., that high frequency modulation experiments provide a characterization of transport phenomena that is intrinsically different with respect to power balance analysis of a stationary phase. It is particularly useful to analyse power switch on/off events and whenever high frequency modulations are not technically feasible. Low-frequency (1-2 Hz) ECRH modulation experiments have been performed on Tore Supra. A large jump (a factor of 8) in the heat diffusivity has been clearly identified at the ECRH power deposition layer. The amplitude and phase of several harmonics of the Fourier transform of the modulated temperature, as well as the time evolution of the modulated temperature have been reproduced by the analytical solution. The jump is found to be much weaker at lower ECRH power (one gyrotron)

Influence of geologic layering on heat transport and storage in an aquifer thermal energy storage system

Science.gov (United States)

Bridger, D. W.; Allen, D. M.

2014-01-01

A modeling study was carried out to evaluate the influence of aquifer heterogeneity, as represented by geologic layering, on heat transport and storage in an aquifer thermal energy storage (ATES) system in Agassiz, British Columbia, Canada. Two 3D heat transport models were developed and calibrated using the flow and heat transport code FEFLOW including: a "non-layered" model domain with homogeneous hydraulic and thermal properties; and, a "layered" model domain with variable hydraulic and thermal properties assigned to discrete geological units to represent aquifer heterogeneity. The base model (non-layered) shows limited sensitivity for the ranges of all thermal and hydraulic properties expected at the site; the model is most sensitive to vertical anisotropy and hydraulic gradient. Simulated and observed temperatures within the wells reflect a combination of screen placement and layering, with inconsistencies largely explained by the lateral continuity of high permeability layers represented in the model. Simulation of heat injection, storage and recovery show preferential transport along high permeability layers, resulting in longitudinal plume distortion, and overall higher short-term storage efficiencies.

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.)

Advances in the optimisation of apparel heating products: A numerical approach to study heat transport through a blanket with an embedded smart heating system

International Nuclear Information System (INIS)

Neves, S.F.; Couto, S.; Campos, J.B.L.M.; Mayor, T.S.

2015-01-01

The optimisation of the performance of products with smart/active functionalities (e. g. in protective clothing, home textiles products, automotive seats, etc.) is still a challenge for manufacturers and developers. The aim of this study was to optimise the thermal performance of a heating product by a numerical approach, by analysing several opposing requirements and defining solutions for the identified limitations, before the construction of the first prototype. A transfer model was developed to investigate the transport of heat from the skin to the environment, across a heating blanket with an embedded smart heating system. Several parameters of the textile material and of the heating system were studied, in order to optimise the thermal performance of the heating blanket. Focus was put on the effects of thickness and thermal conductivity of each layer, and on parameters associated with the heating elements, e.g. position of the heating wires relative to the skin, distance between heating wires, applied heating power, and temperature range for operation of the heating system. Furthermore, several configurations of the blanket (and corresponding heating powers) were analysed in order to minimise the heat loss from the body to the environment, and the temperature distribution along the skin. The results show that, to ensure an optimal compromise between the thermal performance of the product and the temperature oscillation along its surface, the distance between the wires should be small (and not bigger than 50 mm), and each layer of the heating blanket should have a specific thermal resistance, based on the expected external conditions during use and the requirements of the heating system (i.e. requirements regarding energy consumption/efficiency and capacity to effectively regulate body exchanges with surrounding environment). The heating system should operate in an ON/OFF mode based on the body heating needs and within a temperature range specified based on

Diffusive-to-ballistic transition of the modulated heat transport in a rarefied air chamber

Science.gov (United States)

Gomez-Heredia, C. L.; Macias, J.; Ordonez-Miranda, J.; Ares, O.; Alvarado-Gil, J. J.

2017-01-01

Modulated heat transfer in air subject to pressures from 760 Torr to 10-4 Torr is experimentally studied by means of a thermal-wave resonant cavity placed in a vacuum chamber. This is done through the analysis of the amplitude and phase delay of the photothermal signal as a function of the cavity length and pressure through of the Knudsen's number. The viscous, transitional, and free molecular regimes of heat transport are observed for pressures P>1.5 Torr, 25 mTorrheat transport.

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

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)

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

Study of heat transfer and particle transport in Tore Supra and HL-2A tokamaks

International Nuclear Information System (INIS)

Song, S.

2011-12-01

This thesis reports on experimental studies of heat and particles transport performed on 2 large tokamaks: Tore Supra (based at CEA/Cadarache, France) and HL-2A (based at the Southwestern Institute of Physics, Chengdu, China). The modulated source is the Electron Cyclotron Resonance Heating (ECRH) for the heat pinch and density pump-out studies, while the non-local transport experiments use the Supersonic Molecular Beam Injection (SMBI) as source of modulation. The emphasis is put on the inward heat pinch. In the off-axis ECRH modulation experiments on Tore Supra with low frequency (1 Hz), strong heat inward transport has been observed, in particular for low density. Two transport models have been applied in order to analyze the experimental behavior. The first one is the linear pinch model (LPM) and the second one is an empirical model based on micro-instabilities theory, named Critical Gradient Model (CGM). Good agreement has been found for all harmonics between the experimental data and the simulation using LPM. On the other hand, good agreement has not been achieved using CGM. The density pump-out with large particles and energy losses during ECRH is commonly observed in tokamaks. A new dynamic approach using the modulation technique has been used in HL-2A for analyzing the transient phase of the density pump-out. A correlation between the turbulence increase and the density pump-out has been found. The non-local transport phenomenon, characterized by a fast transient process compared to the normal diffusive response to the perturbation is observed. Both phenomena, i.e., pump-out and non-locality, show as simultaneous variation of density and temperature. This can be an inspiration for the usage of a transport matrix which considers the density and temperature evolution together. Simulations with a simple transport matrix, with non-diagonal terms coupling temperature and density qualitatively reproduce the non-local and pump-out effects qualitatively

Skylab and solar exploration. [chromosphere-corona structure, energy production and heat transport processes

Science.gov (United States)

Von Puttkamer, J.

1973-01-01

Review of some of the findings concerning solar structure, energy production, and heat transport obtained with the aid of the manned Skylab space station observatory launched on May 14, 1973. Among the topics discussed are the observation of thermonuclear fusion processes which cannot be simulated on earth, the observation of short-wave solar radiation not visible to observers on earth, and the investigation of energy-transport processes occurring in the photosphere, chromosphere, and corona. An apparent paradox is noted in that the cooler chromosphere is heating the hotter corona, seemingly in defiance of the second law of thermodynamics, thus suggesting that a nonthermal mechanism underlies the energy transport. Understanding of this nonthermal mechanism is regarded as an indispensable prerequisite for future development of plasma systems for terrestrial applications.

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Fee, G.G.

1976-02-01

Brief highlights are presented for the following activities: heavy section steel technology program, fission product β and γ energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, and design criteria for piping and nozzles

Validation study of the COBRA-WC computer program for LMFBR core thermal-hydraulic analysis

International Nuclear Information System (INIS)

Khan, E.U.; Bates, J.M.

1982-01-01

The COBRA-WC (Whole Core) computer program has been developed as a benchmark code to predict flow and temperature fields in LMFBR rod bundles. Consequently, an extensive validation study has been conducted to reinforce its credibility. A set of generalized parameters predicts data well for a wide range of geometries and operating conditions which include conventional (current generation LMFBRs) fuel and blanket assembly geometry in the forced, mixed, and natural convection regimes. The data base used for validating COBRA-WC was obtained from out-of-pile and in-pile tests. Most of the data was obtained in fully heated bundles with bundle power skew across flats up to 3:1 (max:min), Reynolds number between 500 and 80,000, and coolant mixed-mean temperature rise (δ anti T) in the range, 78 0 F less than or equal to δ anti T less than or equal to 340 0 F. Within the bundle, 95% of the predicted coolant temperature data points fall within +-25 0 F for 150 0 F less than or equal to δ anti T less than or equal to 340 0 F and within +-17 0 F for 78 0 F less than or equal to δ anti T less than or equal to 150 0 F

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

Heat and fission product transport in molten core material pool with crust

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2005-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

Status of liquid metal cooled fast breeder reactors

International Nuclear Information System (INIS)

1985-01-01

This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

Transport properties and specific heat of UTe and USb

International Nuclear Information System (INIS)

Ochiai, A.; Suzuki, Y.; Shikama, T.; Suzuki, K.; Hotta, E.; Haga, Y.; Suzuki, T.

1994-01-01

Uranium monochalcogenides and monopnictides crystallize in the NaCl-type structure and exhibit ferromagnetic and antiferromagnetic order, respectively. These series reveal interesting properties such as Kondo behavior of UTe. However, such interesting properties are much sample dependent. We grew single crystals of USb and UTe with high purity using the Bridgman technique, and measured transport properties and specific heat. ((orig.))

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

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)

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

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

Study of Heat Flux Threshold and Perturbation Effect on Transport Barrier Formation Based on Bifurcation Model

International Nuclear Information System (INIS)

Chatthong, B.; Onjun, T.; Imbeaux, F.; Sarazin, Y.; Strugarek, A.; Picha, R.; Poolyarat, N.

2011-06-01

Full text: Formation of transport barrier in fusion plasma is studied using a simple one-field bistable S-curve bifurcation model. This model is characterized by an S-line with two stable branches corresponding to the low (L) and high (H) confinement modes, connected by an unstable branch. Assumptions used in this model are such that the reduction in anomalous transport is caused by v E velocity shear effect and also this velocity shear is proportional to pressure gradient. In this study, analytical and numerical approaches are used to obtain necessary conditions for transport barrier formation, i.e. the ratio of anomalous over neoclassical coefficients and heat flux thresholds which must be exceeded. Several profiles of heat sources are considered in this work including constant, Gaussian, and hyperbolic tangent forms. Moreover, the effect of perturbation in heat flux is investigated with respect to transport barrier formation

Minimization of transport and distribution cost for district heating study of particular cases

International Nuclear Information System (INIS)

Barreau, A.; Caizergues, R.; Moret Bailly, J.

1977-01-01

The transport and distribution of hot pressurized water involve different sets of criteria: transport networks, heat distribution networks, storages. The minimization of transport cost is studied together with the distribution of thermal energy. The same parameters are introduced into these programs. The same method is used for rate of flow calculations, but mathematical methods of pipe diameter calculation are different. Some transport and distribution networks are studied with the corresponding computed programs: 52 branches networks-27 terminations; 287 branches networks-148 terminations

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

3D modeling of groundwater heat transport in the shallow Westliches Leibnitzer Feld aquifer, Austria

Science.gov (United States)

Rock, Gerhard; Kupfersberger, Hans

2018-02-01

For the shallow Westliches Leibnitzer feld aquifer (45 km2) we applied the recently developed methodology by Kupfersberger et al. (2017a) to derive the thermal upper boundary for a 3D heat transport model from observed air temperatures. We distinguished between land uses of grass and agriculture, sealed surfaces, forest and water bodies. To represent the heat flux from heated buildings and the mixture between different land surfaces in urban areas we ran the 1D vertical heat conduction module SoilTemp which is coupled to the heat transport model (using FEFLOW) on a time step basis. Over a simulation period of 23 years the comparison between measured and observed groundwater temperatures yielded NSE values ranging from 0.41 to 0.92 including readings at different depths. The model results showed that the thermal input signals lead to distinctly different vertical groundwater temperature distributions. To overcome the influence of specific warm or cold years we introduced the computation of an annual averaged groundwater temperature profile. With respect to the use of groundwater cooling or heating facilities we evaluated the application of vertically averaged statistical groundwater temperature distributions compared to the use of temperature distributions at selected dates. We concluded that the heat transport model serves well as an aquifer scale management tool to optimize the use of the shallow subsurface for thermal purposes and to analyze the impacts of corresponding measures on groundwater temperatures.

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)

A pumped, two-phase flow heat transport system for orbiting instrument payloads

Science.gov (United States)

Fowle, A. A.

1981-01-01

A pumped two-phase (heat absorption/heat rejection) thermal transport system for orbiting instrument payloads is investigated. The thermofluid characteristics necessary for the system design are discussed. A preliminary design with a series arrangement of four instrument heat stations and six radiators in a single loop is described in detail, and the total mass is estimated to be 134 kg, with the radiators, instrument heat stations, and fluid reservoir accounting for approximately 86, 24, and 12 kg, respectively. The evaluation of preliminary test results shows that the system has potential advantages; however, further research is necessary in the areas of one-g and zero-g heat transfer coefficients/fluid regimes, fluid by-pass temperature control, and reliability of small pumps.

Conceptual design of heat transport systems and components of PFBR-NSSS

International Nuclear Information System (INIS)

Chetal, S.C.; Bhoje, S.B.; Kale, R.D.; Rao, A.S.L.K.; Mitra, T.K.; Selvaraj, A.; Sethi, V.K.; Sundaramoorthy, T.R.; Balasubramaniyan, V.; Vaidyanathan, G.

1996-01-01

The production of electrical power from sodium cooled fast reactors in the present power scenario in India demands emphasis on plant economics consistent with safety. Number of heat transport systems/components and the design of principal heat transport components viz sodium pumps, IHX and steam generators play significant role in the plant capital cost and capacity factor. The paper discusses the basis of selection of 2 primary pumps, 4 IHX, 2 secondary loops, 2 secondary pumps and 8 steam generators for the 500 MWe Prototype Fast Breeder Reactor (PFBR), which is now in design stage. The principal design features of primary pump, IHX and steam generator have been selected based on design simplicity, ease of manufacture and utilization of established designs. The paper also describes the conceptual design of above mentioned three components. (author). 3 figs, 2 tabs

An alternative treatment of heat flow for charge transport in semiconductor devices

International Nuclear Information System (INIS)

Grupen, Matt

2009-01-01

A unique thermodynamic model of Fermi gases suitable for semiconductor device simulation is presented. Like other models, such as drift diffusion and hydrodynamics, it employs moments of the Boltzmann transport equation derived using the Fermi-Dirac distribution function. However, unlike other approaches, it replaces the concept of an electron thermal conductivity with the heat capacity of an ideal Fermi gas to determine heat flow. The model is used to simulate a field-effect transistor and show that the external current-voltage characteristics are strong functions of the state space available to the heated Fermi distribution.

Bouyancy effects on sodium coolant temperature profiles measured in an electrically heated mock-up of a 61-rod breeder reactor blanket assembly

International Nuclear Information System (INIS)

Engel, F.C.; Markley, R.A.; Minushkin, B.

1978-01-01

The paper describes test results selected to demonstrate the effect of buoyancy on the temperature profiles in a 61-rod electrically heated mock-up of an LMFBR radial blanket assembly. In these assemblies, heat transfer occurs over a wide range of complex operating conditions. The range and complexity of conditions are the result of the steep flux and power gradients which are an inherent feature of the blanket region and the power generation level in an assembly which can vary from 20 to 1100 kW

Improved inter-assembly heat transfer modeling under low flow conditions for the Super System Code (SSC)

International Nuclear Information System (INIS)

Horak, W.C.; Guppy, J.G.

1984-01-01

The Super System Code (SSC) was developed at the Brookhaven National Laboratory (BNL) for the thermal hydraulic analysis of natural circulation transients, operational transients, and other system wide transients in nuclear power plants. SSC is a generic, best estimate code that models the in-vessel components, heat transport loops, plant protection systems and plant control systems. SSC also simulates the balance of plant when interfaced with the MINET code. SSC has been validated against both numerical and experimental data bases and is now used by several outside users. An important area of interest in LMFBR transient analysis is the prediction of the response of the reactor core under low flow conditions, such as experienced during a natural circulation event. Under these circumstances there are many physical phenomena which must be modeled to provide an adequate representation by a computer code simulation. The present version of SSC contains numerous models which account for most of the major phenomena. However, one area where the present model in SSC is being improved is in the representation of heat transfer and buoyancy effects under low flow operation. To properly improve the present version, the addition of models to represent certain inter-assembly effects is required

Increased Heat Transport in Ultra-hot Jupiter Atmospheres through H2 Dissociation and Recombination

Science.gov (United States)

Bell, Taylor J.; Cowan, Nicolas B.

2018-04-01

A new class of exoplanets is beginning to emerge: planets with dayside atmospheres that resemble stellar atmospheres as most of their molecular constituents dissociate. The effects of the dissociation of these species will be varied and must be carefully accounted for. Here we take the first steps toward understanding the consequences of dissociation and recombination of molecular hydrogen (H2) on atmospheric heat recirculation. Using a simple energy balance model with eastward winds, we demonstrate that H2 dissociation/recombination can significantly increase the day–night heat transport on ultra-hot Jupiters (UHJs): gas giant exoplanets where significant H2 dissociation occurs. The atomic hydrogen from the highly irradiated daysides of UHJs will transport some of the energy deposited on the dayside toward the nightside of the planet where the H atoms recombine into H2; this mechanism bears similarities to latent heat. Given a fixed wind speed, this will act to increase the heat recirculation efficiency; alternatively, a measured heat recirculation efficiency will require slower wind speeds after accounting for H2 dissociation/recombination.

Design to nullify activity movement in heat transport systems

International Nuclear Information System (INIS)

Hemmings, R.L.; Barber, D.

1975-01-01

This article describes the methods by which designers can reduce the adverse effects of system corrosion and the resultant activation of the corrosion products in heat transport systems. The presentation will cover: a) choice of materials; b) assessment of the need of components; c) control of system chemistry; d) factors considered in sizing HTS purification systems; i) control of activation and fission products; ii) decontamination. (author)

FFTF Heat Transport System (HTS) component and system design

International Nuclear Information System (INIS)

Young, M.W.; Edwards, P.A.

1980-01-01

The FFTF Heat Transport Systems and Components designs have been completed and successfully tested at isothermal conditions up to 427 0 C (800 0 F). General performance has been as predicted in the design analyses. Operational flexibility and reliability have been outstanding throughout the test program. The components and systems have been demonstrated ready to support reactor powered operation testing planned later in 1980

Comparing the value of bioenergy in the heating and transport sectors of an electricity-intensive energy system in Norway

International Nuclear Information System (INIS)

Assefa Hagos, Dejene; Gebremedhin, Alemayehu; Folsland Bolkesjø, Torjus

2015-01-01

The objective of this paper is to identify the most valuable sector for the use of bioenergy in a flexible energy system in order to meet the energy policy objectives of Inland Norway. A reference system was used to construct alternative systems in the heating and transport sectors. The alternative system in the heating sector is based on heat pumps and bio-heat boilers while the alternative systems in the transport sector are based on three different pathways: bio-dimethyl ether, hydrogen fuel cell vehicles and battery electric vehicles. The alternative systems were compared with the reference system after a business-economic optimisation had been made using an energy system analysis tool. The results show that the excess electricity availability due to increased energy efficiency measures hampers the competitiveness and penetration of bio-heating over heat pumps in the heating sector. Indeed, the synergy effect of using bio-dimethyl ether in the transport sector for an increased share of renewable energy sources is much higher than that of the hydrogen fuel cell vehicle and battery electric vehicle pathways. The study also revealed that increasing renewable energy production would increase the renewable energy share more than what would be achieved by an increase in energy efficiency. -- Highlights: •Bio-heating is less competitive over heat pump for low quality heat production. •Renewable energy production meets policy objectives better than system efficiency. •Bioenergy is more valuable in the transport sector than the heating sector

Heat transport inventory monitoring for CANDU-PHW reactors

International Nuclear Information System (INIS)

Hussein, E.; Luxat, J.C.

1984-01-01

A computer-based D 2 O coolant inventory monitoring system proposed for implementation on the digital computer controllers at Ontario Hydro's CANDU generating units is discussed. By monitoring process parameters and utilizing probabilistically-based decision algorithms, timely indication of any significant loss of D 2 O inventory will be provided to the operator. The monitoring is performed in a co-ordinated manner such that D 2 O losses from either the heat transport system or the inventory control system can be detected. (orig.)

Tokamak electron heat transport by direct numerical simulation of small scale turbulence

International Nuclear Information System (INIS)

Labit, B.

2002-10-01

In a fusion machine, understanding plasma turbulence, which causes a degradation of the measured energy confinement time, would constitute a major progress in this field. In tokamaks, the measured ion and electron thermal conductivities are of comparable magnitude. The possible sources of turbulence are the temperature and density gradients occurring in a fusion plasma. Whereas the heat losses in the ion channel are reasonably well understood, the origin of the electron losses is more uncertain. In addition to the radial velocity associated to the fluctuations of the electric field, electrons are more affected than ions by the magnetic field fluctuations. In experiments, the confinement time can be conveniently expressed in terms of dimensionless parameters. Although still somewhat too imprecise, these scaling laws exhibit strong dependencies on the normalized pressure β or the normalized Larmor radius, ρ * . The present thesis assesses whether a tridimensional, electromagnetic, nonlinear fluid model of plasma turbulence driven by a specific instability can reproduce the dependence of the experimental electron heat losses on the dimensionless parameters β and ρ * . The investigated interchange instability is the Electron Temperature Gradient driven one (ETG). The model is built by using the set of Braginskii equations. The developed simulation code is global in the sense that a fixed heat flux is imposed at the inner boundary, leaving the gradients free to evolve. From the nonlinear simulations, we have put in light three characteristics for the ETG turbulence: the turbulent transport is essentially electrostatic; the potential and pressure fluctuations form radially elongated cells called streamers; the transport level is very low compared to the experimental values. The thermal transport dependence study has shown a very small role of the normalized pressure, which is in contradiction with the Ohkama's formula. On the other hand, the crucial role of the

Influence of transport on EBW heating efficiency in magnetic confinement devices

International Nuclear Information System (INIS)

Cappa, A.; Castejon, F.; Lopez-Bruna, D.; Tereshchenko, M.

2007-01-01

The main advantage of the heating performed by electron Bernstein waves (EBW) in the O-X-B1 regime (O mode injection that is converted into X mode, which is converted in Bernstein wave, strongly absorbed close to the cyclotron resonance layer at first harmonic) is that there is no cut-off density. Therefore, this heating system can work without upper density limit, still having all the advantages of electron cyclotron resonance heating (ECRH), which is localised in phase space due to its resonant nature. The heating efficiency of Bernstein waves depends on the fraction of waves that is transformed from O to X mode at the O mode cut off layer, then on the fraction of power converted into Bernstein waves at the upper hybrid resonance layer and, finally, on the final position of the absorption in the plasma. All these factors are related to the density profile, since the positions of the cut off and of the upper hybrid resonance layers depend on the actual plasma density profile. Besides, the absorption profile depends also on the temperature profile. Moreover, it is possible to observe that the former layers only appear for high enough plasma density, than can be obtained by gas puffing, as has been observed in the simulations performed for TJ-II stellarator. For such reasons, particle transport is basic for understanding and guaranteeing EBW heating. In this work, TJ-II plasmas are taken as a case example in order to simulate the full evolution of a plasma discharge that is created and heated by ECRH in a first step and finally is heated using EBW. The evolution of the discharge is simulated using the transport code ASTRA and the sequence of the discharge is as follows: O mode is launched on a steady state plasma with density lower than the O mode cut-off. Then a gas puff is injected in order to increase the plasma density over the level in which EBW heating is efficient because O mode cut off and upper hybrid layer appear. EBW ray tracing calculations are performed

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

Local and Nonlocal Parallel Heat Transport in General Magnetic Fields

International Nuclear Information System (INIS)

Castillo-Negrete, D. del; Chacon, L.

2011-01-01

A novel approach for the study of parallel transport in magnetized plasmas is presented. The method avoids numerical pollution issues of grid-based formulations and applies to integrable and chaotic magnetic fields with local or nonlocal parallel closures. In weakly chaotic fields, the method gives the fractal structure of the devil's staircase radial temperature profile. In fully chaotic fields, the temperature exhibits self-similar spatiotemporal evolution with a stretched-exponential scaling function for local closures and an algebraically decaying one for nonlocal closures. It is shown that, for both closures, the effective radial heat transport is incompatible with the quasilinear diffusion model.

Thermal performance of the nuclear fuel rods submitted to angular variation of the heat exchanger coefficients

International Nuclear Information System (INIS)

Carvalho, A.M.M. de.

1984-01-01

Generally, LMFBR fuel rods consist of fuel pellets encapsulated in cladding tubes. These tubes are wrapped by a helical wire, working as a spacer. Distortions in the rod temperature distribution and in the external heat flux can be generated by angular variations in the local heat transfer coefficients due to the wire, by excentricity between pellet and clad or by ovalization of the cladding tube. Also, the temperature distributions can be affected by fuel densification, reestructuring and swelling. The present work consists of the development of a computer code in order to analyse the fuel rod performance as function of geometrical and operational effects, in steady state regime. (Author) [pt

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.)

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Fee, G.G.

1976-08-01

Brief highlights are presented for the following programs: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, multirod burst tests, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis, design criteria for piping and nozzles, and dose conversion factors for inhalation of radionuclides

Monthly highlights for Office of Nuclear Regulatory Research Programs at Oak Ridge National Laboratory, August 1976

International Nuclear Information System (INIS)

Fee, G.G.

1976-10-01

Technical highlights are presented for the following activities: heavy section steel technology, fission product beta and gamma energy release, LOCA release from LWR fuel, Nuclear Safety Information Center, PWR blowdown heat transfer-separate effects, Zircaloy fuel cladding collapse studies, zirconium metal-water oxidation kinetics, aerosol release and transport from LMFBR fuel, HTGR safety analysis and research, design criteria for piping and nozzles, and dose conversion factors for inhalation of radionuclides

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

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.)

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.)

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)

Exact harmonic solutions to Guyer-Krumhansl-type equation and application to heat transport in thin films

Science.gov (United States)

Zhukovsky, K.; Oskolkov, D.

2018-03-01

A system of hyperbolic-type inhomogeneous differential equations (DE) is considered for non-Fourier heat transfer in thin films. Exact harmonic solutions to Guyer-Krumhansl-type heat equation and to the system of inhomogeneous DE are obtained in Cauchy- and Dirichlet-type conditions. The contribution of the ballistic-type heat transport, of the Cattaneo heat waves and of the Fourier heat diffusion is discussed and compared with each other in various conditions. The application of the study to the ballistic heat transport in thin films is performed. Rapid evolution of the ballistic quasi-temperature component in low-dimensional systems is elucidated and compared with slow evolution of its diffusive counterpart. The effect of the ballistic quasi-temperature component on the evolution of the complete quasi-temperature is explored. In this context, the influence of the Knudsen number and of Cauchy- and Dirichlet-type conditions on the evolution of the temperature distribution is explored. The comparative analysis of the obtained solutions is performed.

An Assessment of Transport Property Estimation Methods for Ammonia–Water Mixtures and Their Influence on Heat Exchanger Size

DEFF Research Database (Denmark)

Kærn, Martin Ryhl; Modi, Anish; Jensen, Jonas Kjær

2015-01-01

Transport properties of fluids are indispensable for heat exchanger design. The methods for estimating the transport properties of ammonia–water mixtures are not well established in the literature. The few existent methods are developed from none or limited, sometimes inconsistent experimental...... of ammonia–water mixtures. Firstly, the different methods are introduced and compared at various temperatures and pressures. Secondly, their individual influence on the required heat exchanger size (surface area) is investigated. For this purpose, two case studies related to the use of the Kalina cycle...... the interpolative methods in contrast to the corresponding state methods. Nevertheless, all possible mixture transport property combinations used herein resulted in a heat exchanger size within 4.3 % difference for the flue-gas heat recovery boiler, and within 12.3 % difference for the oil-based boiler....

Relationship between particle and heat transport in JT-60U plasmas with internal transport barrier

International Nuclear Information System (INIS)

Takenaga, Hidenobu; Higashijima, S.; Oyama, N.

2003-01-01

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high β p ELMy H-mode plasmas in JT-60U. No helium and carbon accumulation inside the ITB is observed even with ion heat transport reduced to a neoclassical level. On the other hand, the heavy impurity argon is accumulated inside the ITB. The argon density profile estimated from the soft x-ray profile is more peaked, by a factor of 2-4 in the RS plasma and of 1.6 in the high β p mode plasma, than the electron density profile. The helium diffusivity (D He ) and the ion thermal diffusivity (χ i ) are at an anomalous level in the high β p mode plasma, where D He and χ i are higher by a factor of 5-10 than the neoclassical value. In the RS plasma, D He is reduced from the anomalous to the neoclassical level, together with χ i . The carbon and argon density profiles calculated using the transport coefficients reduced to the neoclassical level only in the ITB are more peaked than the measured profiles, even when χ i is reduced to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying ECH in the high β p mode plasma, where both electron and argon density profiles become flatter. The reduction of the neoclassical inward velocity for argon due to the reduction of density gradient is consistent with the experimental observation. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density control to suppress heavy impurity accumulation. (author)

Relationship between particle and heat transport in JT-60U plasmas with internal transport barrier

International Nuclear Information System (INIS)

Takenaga, H.; Higashijima, S.; Oyama, N.

2003-01-01

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high β p ELMy H-mode plasmas in JT-60U. No helium and carbon accumulation inside the ITB is observed even with ion heat transport reduced to a neoclassical level. On the other hand, the heavy impurity argon is accumulated inside the ITB. The argon density profile estimated from the soft x-ray profile is more peaked, by a factor of 2-4 in the RS plasma and of 1.6 in the high β p mode plasma, than the electron density profile. The helium diffusivity (D He ) and the ion thermal diffusivity (χ i ) are at an anomalous level in the high β p mode plasma, where D He and χ i are higher by a factor of 5-10 than the neoclassical value. In the RS plasma, D He is reduced from the anomalous to the neoclassical level, together with χ i . The carbon and argon density profiles calculated using the transport coefficients reduced to the neoclassical level only in the ITB are more peaked than the measured profiles, even when χ i is reduced to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying ECH in the high β p mode plasma, where both electron and argon density profiles become flatter. The reduction of the neoclassical inward velocity for argon due to the reduction of density gradient is consistent with the experimental observation. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density gradient control to suppress heavy impurity accumulation. (author)

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)

Heat pulse analysis in JET and relation to local energy transport models

International Nuclear Information System (INIS)

Haas, J.C.M. de; Lopes Cardozo, N.J.; Han, W.; Sack, C.; Taroni, A.

1989-01-01

The evolution of a perturbation T e of the electron temperature depends on the linearised expression of the heat flux q e and may be not simply related to the local value of the electron heat conductivity χ e . It is possible that local heat transport models predicting similar temperature profiles and global energy confinement properties, imply a different propagation of heat pulses. We investigate here this possibility for the case of two models developed at JET. We also present results obtained at JET on a set of discharges covering the range of currents from 2 to 5 MA. Only L-modes, limiter discharges are considered here. Experimental results on the scaling of χ HP , the value of χ e related to heat pulse propagation, are compared with those of χ HP derived from the models. (author) 7 refs., 2 figs., 2 tabs

Heat transport and afterheat removal for gas cooled reactors under accident conditions

International Nuclear Information System (INIS)

2001-01-01

The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

Analysis of simulation methodology for calculation of the heat of transport for vacancy thermodiffusion

Energy Technology Data Exchange (ETDEWEB)

Tucker, William C.; Schelling, Patrick K., E-mail: patrick.schelling@ucf.edu [Advanced Material Processing and Analysis Center and Department of Physics, University of Central Florida, 4000 Central Florida Blvd., Orlando, Florida 32816 (United States)

2014-07-14

Computation of the heat of transport Q{sub a}{sup *} in monatomic crystalline solids is investigated using the methodology first developed by Gillan [J. Phys. C: Solid State Phys. 11, 4469 (1978)] and further developed by Grout and coworkers [Philos. Mag. Lett. 74, 217 (1996)], referred to as the Grout-Gillan method. In the case of pair potentials, the hopping of a vacancy results in a heat wave that persists for up to 10 ps, consistent with previous studies. This leads to generally positive values for Q{sub a}{sup *} which can be quite large and are strongly dependent on the specific details of the pair potential. By contrast, when the interactions are described using the embedded atom model, there is no evidence of a heat wave, and Q{sub a}{sup *} is found to be negative. This demonstrates that the dynamics of vacancy hopping depends strongly on the details of the empirical potential. However, the results obtained here are in strong disagreement with experiment. Arguments are presented which demonstrate that there is a fundamental error made in the Grout-Gillan method due to the fact that the ensemble of states only includes successful atom hops and hence does not represent an equilibrium ensemble. This places the interpretation of the quantity computed in the Grout-Gillan method as the heat of transport in doubt. It is demonstrated that trajectories which do not yield hopping events are nevertheless relevant to computation of the heat of transport Q{sub a}{sup *}.

Thermal hydraulic characteristics of the pool-type FBR 'ARES' eliminating intermediate heat-transfer system. 2. Transient characteristics in representative events

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Ueda, Nobuyuki; Kinoshita, Izumi; Yoshida, Kazuo

1999-01-01

An innovative plant concept (ARES), which eliminates intermediate heat-transfer system, was proposed in order to reduce the construction cost of the liquid metal cooled fast breeder reactor (LMFBR). In the present study, representative events of the ARES were picked up and plant transient analysis was conducted in order to reveal the thermal hydraulic characteristics of the ARES. The main results of the analysis are as follows: (1) The core was sufficiently cooled without sodium boiling in an event of mis-operation of the siphon break system. (2) The disturbances of the SG transfer to the core within about 30 seconds in a event of the feed water increasing. This transition period is less than half that of the conventional LMFBR. However, the transient characteristics of the ARES are very similar to those of the conventional LMFBR. Based on these results, it is only necessary to consider the short transition period in order to design the control system. (3) Based on the results of the calculation of a total blackout accident, it is concluded that the use of the SG-side cooling together is profitable in order to satisfy the design safety criteria in design basis events. In the case of full natural-circulation cooling, the use of the SG-side cooling together is profitable in order to satisfy the design safety criteria. (author)

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

Transient heat transport studies in JET conventional and advanced tokamak plasmas

International Nuclear Information System (INIS)

Mantica, P.; Coffey, I.; Dux, R.

2003-01-01

Transient transport studies are a valuable complement to steady-state analysis for the understanding of transport mechanisms and the validation of physics-based transport models. This paper presents results from transient heat transport experiments in JET and their modelling. Edge cold pulses and modulation of ICRH (in mode conversion scheme) have been used to provide detectable electron and ion temperature perturbations. The experiments have been performed in conventional L-mode plasmas or in Advanced Tokamak regimes, in the presence of an Internal Transport Barrier (ITB). In conventional plasmas, the issues of stiffness and non-locality have been addressed. Cold pulse propagation in ITB plasmas has provided useful insight into the physics of ITB formation. The use of edge perturbations for ITB triggering has been explored. Modelling of the experimental results has been performed using both empirical models and physics-based models. Results of cold pulse experiments in ITBs have also been compared with turbulence simulations. (author)

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.)

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

Acoustically enhanced heat transport

Energy Technology Data Exchange (ETDEWEB)

Ang, Kar M.; Hung, Yew Mun; Tan, Ming K., E-mail: tan.ming.kwang@monash.edu [School of Engineering, Monash University Malaysia, 47500 Bandar Sunway, Selangor (Malaysia); Yeo, Leslie Y. [Micro/Nanophysics Research Laboratory, RMIT University, Melbourne, VIC 3001 (Australia); Friend, James R. [Department of Mechanical and Aerospace Engineering, University of California, San Diego, California 92093 (United States)

2016-01-15

We investigate the enhancement of heat transfer in the nucleate boiling regime by inducing high frequency acoustic waves (f ∼ 10{sup 6} Hz) on the heated surface. In the experiments, liquid droplets (deionized water) are dispensed directly onto a heated, vibrating substrate. At lower vibration amplitudes (ξ{sub s} ∼ 10{sup −9} m), the improved heat transfer is mainly due to the detachment of vapor bubbles from the heated surface and the induced thermal mixing. Upon increasing the vibration amplitude (ξ{sub s} ∼ 10{sup −8} m), the heat transfer becomes more substantial due to the rapid bursting of vapor bubbles happening at the liquid-air interface as a consequence of capillary waves travelling in the thin liquid film between the vapor bubble and the air. Further increases then lead to rapid atomization that continues to enhance the heat transfer. An acoustic wave displacement amplitude on the order of 10{sup −8} m with 10{sup 6} Hz order frequencies is observed to produce an improvement of up to 50% reduction in the surface temperature over the case without acoustic excitation.

Heat and momentum transport scalings in vertical convection

Science.gov (United States)

Shishkina, Olga

2016-11-01

For vertical convection, where a fluid is confined between two differently heated isothermal vertical walls, we investigate the heat and momentum transport, which are measured, respectively, by the Nusselt number Nu and the Reynolds number Re . For laminar vertical convection we derive analytically the dependence of Re and Nu on the Rayleigh number Ra and the Prandtl number Pr from our boundary layer equations and find two different scaling regimes: Nu Pr 1 / 4 Ra 1 / 4 , Re Pr - 1 / 2 Ra 1 / 2 for Pr > 1 . Direct numerical simulations for Ra from 105 to 1010 and Pr from 0.01 to 30 are in excellent ageement with our theoretical findings and show that the transition between the regimes takes place for Pr around 0.1. We summarize the results from and present new theoretical and numerical results for transitional and turbulent vertical convection. The work is supported by the Deutsche Forschungsgemeinschaft (DFG) under the Grant Sh 405/4 - Heisenberg fellowship.

Thermal relaxation and heat transport in spin ice Dy{sub 2}Ti{sub 2}O{sub 7}

Energy Technology Data Exchange (ETDEWEB)

Klemke, Bastian; Meissner, M.; Tennant, D.A. [Helmholtz-Zentrum Berlin (Germany); Technische Universitaet Berlin (Germany); Strehlow, P. [Technische Universitaet Berlin (Germany); Physikalisch Technische Bundesanstalt, Institut Berlin (Germany); Kiefer, K. [Helmholtz-Zentrum Berlin (Germany); Grigera, S.A. [School of Physics and Astronomy, St. Andrews (United Kingdom); Instituto de Fisica de Liquidos y Sistemas Biologicos, CONICET, UNLP, La Plata (Argentina)

2011-07-01

The thermal properties of single crystalline Dy{sub 2}Ti{sub 2}O{sub 7} have been studied at temperature below 30 K and magnetic fields applied along [110] direction up to 1.5 T. Based on a thermodynamic field theory (TFT) various heat relaxation and thermal transport measurements were analysed. So we were able to present not only the heat capacity of Dy{sub 2}Ti{sub 2}O{sub 7}, but also for the first time the different contributions of the magnetic excitations and their corresponding relaxation times in the spin ice phase. In addition, the thermal conductivity and the shortest relaxation time were determined by thermodynamic analysis of steady state heat transport measurements. Finally, we were able to reproduce the temperature profiles recorded in heat pulse experiments on the basis of TFT using the previously determined heat capacity and thermal conductivity data without additional parameters. Thus, TFT has been proved to be thermodynamically consistent in describing three thermal transport experiments on different time scales. The observed temperature and field dependencies of heat capacity contributions and relaxation times indicate the magnetic excitations in the spin ice Dy{sub 2}Ti{sub 2}O{sub 7} as thermally activated monopole-antimonopole defects.

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

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

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.

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.

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.

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

Two-phase optimizing approach to design assessments of long distance heat transportation for CHP systems

International Nuclear Information System (INIS)

Hirsch, Piotr; Duzinkiewicz, Kazimierz; Grochowski, Michał; Piotrowski, Robert

2016-01-01

Highlights: • New method for long distance heat transportation system effectivity evaluation. • Decision model formulation which reflects time and spatial structure of the problem. • Multi-criteria and complex approach to solving the decision-making problem. • Solver based on simulation-optimization approach with two-phase optimization method. • Sensitivity analysis of the optimization procedure elements. - Abstract: Cogeneration or Combined Heat and Power (CHP) for power plants is a method of putting to use waste heat which would be otherwise released to the environment. This allows the increase in thermodynamic efficiency of the plant and can be a source of environmental friendly heat for District Heating (DH). In the paper CHP for Nuclear Power Plant (NPP) is analyzed with the focus on heat transportation. A method for effectivity and feasibility evaluation of the long distance, high power Heat Transportation System (HTS) between the NPP and the DH network is proposed. As a part of the method the multi-criteria decision-making problem, having the structure of the mathematical programming problem, for optimized selection of design and operating parameters of the HTS is formulated. The constraints for this problem include a static model of HTS, that allows considerations of system lifetime, time variability and spatial topology. Thereby variation of annual heat demand within the DH area, variability of ground temperature, insulation and pipe aging and/or terrain elevation profile can be taken into account in the decision-making process. The HTS construction costs, pumping power, and heat losses are considered as objective functions. In general, the analyzed optimization problem is multi-criteria, hybrid and nonlinear. The two-phase optimization based on optimization-simulation framework is proposed to solve the decision-making problem. The solver introduces a number of assumptions concerning the optimization process. Methods for problem decomposition

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

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

LOA-1: prevent accidents. Quarterly technical progress report, FRSP program - July through September 1981

International Nuclear Information System (INIS)

1981-01-01

Information related to LMFBR reactor safety is presented concerning common cause failures; shutdown by self-activated system; shutdown heat removal system operation; sodium burning; core catcher material interactions; accident release of sodium oxide aerosol; and LMFBR risk assessment

Parallel heat transport in integrable and chaotic magnetic fields

Energy Technology Data Exchange (ETDEWEB)

Castillo-Negrete, D. del; Chacon, L. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-8071 (United States)

2012-05-15

The study of transport in magnetized plasmas is a problem of fundamental interest in controlled fusion, space plasmas, and astrophysics research. Three issues make this problem particularly challenging: (i) The extreme anisotropy between the parallel (i.e., along the magnetic field), {chi}{sub ||} , and the perpendicular, {chi}{sub Up-Tack }, conductivities ({chi}{sub ||} /{chi}{sub Up-Tack} may exceed 10{sup 10} in fusion plasmas); (ii) Nonlocal parallel transport in the limit of small collisionality; and (iii) Magnetic field lines chaos which in general complicates (and may preclude) the construction of magnetic field line coordinates. Motivated by these issues, we present a Lagrangian Green's function method to solve the local and non-local parallel transport equation applicable to integrable and chaotic magnetic fields in arbitrary geometry. The method avoids by construction the numerical pollution issues of grid-based algorithms. The potential of the approach is demonstrated with nontrivial applications to integrable (magnetic island), weakly chaotic (Devil's staircase), and fully chaotic magnetic field configurations. For the latter, numerical solutions of the parallel heat transport equation show that the effective radial transport, with local and non-local parallel closures, is non-diffusive, thus casting doubts on the applicability of quasilinear diffusion descriptions. General conditions for the existence of non-diffusive, multivalued flux-gradient relations in the temperature evolution are derived.

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.

Magnetic flux tubes and transport of heat in the convection zone of the sun

International Nuclear Information System (INIS)

Spruit, H.C.

1977-01-01

This thesis consists of five papers dealing with transport of heat in the solar convection zone on the one hand, and with the structure of magnetic flux tubes in the top of the convection zone on the other hand. These subjects are interrelated. For example, the heat flow in the convection zone is disturbed by the presence of magnetic flux tubes, while exchange of heat between a flux tube and the convection zone is important for the energy balance of such a tube. A major part of this thesis deals with the structure of small magnetic flux tubes. Such small tubes (diameters less than about 2'') carry most of the flux appearing at the solar surface. An attempt is made to construct models of the surface layers of such small tubes in sufficient detail to make a comparison with observations possible. Underlying these model calculations is the assumption that the magnetic elements at the solar surface are flux tubes in a roughly static equilibrium. The structure of such tubes is governed by their pressure equilibrium, exchange of heat with the surroundings, and transport of heat by some modified form of convection along the tube. The tube models calculated are compared with observations